Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)

International Nuclear Information System (INIS)

1980-10-01

Since 1970, solid waste with TRU or U-233 contamination in excess of 10 μCi per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact

Radioactive Waste Management Information for 1992 and record-to-date

International Nuclear Information System (INIS)

Litteer, D.L.; Randall, V.C.; Sims, A.M.; Taylor, K.A.

1993-07-01

This document provides detailed data and graphics on air borne and liquid effluent releases, fuel oil and coal consumption, water usage, and hazardous and mixed waste generated for calendar year 1992. This report summarizes industrial waste data records compiled since 1971 for the Idaho National Engineering Laboratory (INEL). The data presented are from the INEL Nonradiological Waste Management Information System

Impact of radioactive waste management operations

International Nuclear Information System (INIS)

Paine, D.; Rogers, L.E.; Uresk, D.W.

1977-01-01

Impact assessment of radioactive waste management operations is considered separately for nonradiological impact on biota, impact on ecosystem structure and function and radiological impact on biota. Localized effects related to facility construction and maintenance activities probably occur but the large expanse of relatively undisturbed surrounding landscape minimizes any overall effects

Liquid and Gaseous Waste Operations Department annual operating report CY 1996

International Nuclear Information System (INIS)

Maddox, J.J.; Scott, C.B.

1997-03-01

This annual report summarizes operating activities dealing with the process waste system, the liquid low-level waste system, and the gaseous waste system. It also describes upgrade activities dealing with the process and liquid low-level waste systems, the cathodic protection system, a stack ventilation system, and configuration control. Maintenance activities are described dealing with nonradiological wastewater treatment plant, process waste treatment plant and collection system, liquid low-level waste system, and gaseous waste system. Miscellaneous activities include training, audits/reviews/tours, and environmental restoration support

Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination

International Nuclear Information System (INIS)

Jacobson, Victor Levon

2002-01-01

U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodium-bearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements

Non-radiological consequences to the aquatic biota and fisheries of the Susquehanna River from the 1979 accident at Three Mile Island Nuclear Station

International Nuclear Information System (INIS)

Hickey, C.R. Jr.; Samworth, R.B.

1979-11-01

The non-radiological consequences to the aquatic biota and fishes of the Susquehanna River from the March 28, 1979 accident at Three Mile Island Nuclear Station were assessed through the post-accident period of July 1979. Thermal and chemical discharges during the period did not exceed required effluent limitations. Several million gallons of treated industrial waste effluents were released into the river which were not of unusual volumes compared with normal operation and were a very small proportion of the seasonally high river flows. The extent and relative location of the effluent plume were defined and the fisheries known to have been under its immediate influence were identified, including rough, forage, and predator/sport fishery species

Development of a comprehensive radioactive waste classification system

International Nuclear Information System (INIS)

Smith, C.F.; Cohen, J.J.

1989-01-01

Several previous studies have been conducted with the intent of developing a rational system for classification of radioactive wastes. Although none of the proposed systems has gained general acceptance, certain waste classes, specifically high-level waste and low-level waste suitable for shallow land burial have been essentially defined by regulation. Wastes which remain undefined include: those intermediate level wastes which require more restrictive controls than that provided by shallow land burial but not the high degree of isolation needed for high level wastes, and wastes below regulatory concern (BRC) which entail so low a radiological risk that they can be managed according to their nonradiological properties. This study has developed a framework within which the complete spectrum of radioactive wastes can be defined

Risks associated with nuclear material recovery and waste preparation

Energy Technology Data Exchange (ETDEWEB)

Fullwood, R R; Erdmann, R C

1983-01-01

An analysis of the risk associated with nuclear material recovery and waste preparation is presented. The steps involve: reprocessing of spent fuel to recycle fissionable material, refabrication of the recovered material for use as reactor fuel, and the transportation links connecting these plants with the power plants and waste repositories. The risks considered are radiological and non-radiological, accident and routine effects on the public and workers during plant construction, operation and decommissioning.

Transportation risk assessment of radioactive wastes generated by the N-Reactor stabilization program at the Hanford Site, Washington

International Nuclear Information System (INIS)

Wheeler, T.

1994-12-01

The potential radiological and nonradiological risks associated with specific radioactive waste shipping campaigns at the Hanford Site are estimated. The shipping campaigns analyzed are associated with the transportation of wastes from the N-Reactor site at the 200-W Area, both within the Hanford Reservation, for disposal. The analysis is based on waste that would be generated from the N-Reactor stabilization program

Hazards assessment for the Waste Experimental Reduction Facility

International Nuclear Information System (INIS)

Calley, M.B.; Jones, J.L. Jr.

1994-01-01

This report documents the hazards assessment for the Waste Experimental Reduction Facility (WERF) located at the Idaho National Engineering Laboratory, which is operated by EG ampersand G Idaho, Inc., for the US Department of Energy (DOE). The hazards assessment was performed to ensure that this facility complies with DOE and company requirements pertaining to emergency planning and preparedness for operational emergencies. DOE Order 5500.3A requires that a facility-specific hazards assessment be performed to provide the technical basis for facility emergency planning efforts. This hazards assessment was conducted in accordance with DOE Headquarters and DOE Idaho Operations Office (DOE-ID) guidance to comply with DOE Order 5500.3A. The hazards assessment identifies and analyzes hazards that are significant enough to warrant consideration in a facility's operational emergency management program. This hazards assessment describes the WERF, the area surrounding WERF, associated buildings and structures at WERF, and the processes performed at WERF. All radiological and nonradiological hazardous materials stored, used, or produced at WERF were identified and screened. Even though the screening process indicated that the hazardous materials could be screened from further analysis because the inventory of radiological and nonradiological hazardous materials were below the screening thresholds specified by DOE and DOE-ID guidance for DOE Order 5500.3A, the nonradiological hazardous materials were analyzed further because it was felt that the nonradiological hazardous material screening thresholds were too high

Hazards assessment for the Waste Experimental Reduction Facility

Energy Technology Data Exchange (ETDEWEB)

Calley, M.B.; Jones, J.L. Jr.

1994-09-19

This report documents the hazards assessment for the Waste Experimental Reduction Facility (WERF) located at the Idaho National Engineering Laboratory, which is operated by EG&G Idaho, Inc., for the US Department of Energy (DOE). The hazards assessment was performed to ensure that this facility complies with DOE and company requirements pertaining to emergency planning and preparedness for operational emergencies. DOE Order 5500.3A requires that a facility-specific hazards assessment be performed to provide the technical basis for facility emergency planning efforts. This hazards assessment was conducted in accordance with DOE Headquarters and DOE Idaho Operations Office (DOE-ID) guidance to comply with DOE Order 5500.3A. The hazards assessment identifies and analyzes hazards that are significant enough to warrant consideration in a facility`s operational emergency management program. This hazards assessment describes the WERF, the area surrounding WERF, associated buildings and structures at WERF, and the processes performed at WERF. All radiological and nonradiological hazardous materials stored, used, or produced at WERF were identified and screened. Even though the screening process indicated that the hazardous materials could be screened from further analysis because the inventory of radiological and nonradiological hazardous materials were below the screening thresholds specified by DOE and DOE-ID guidance for DOE Order 5500.3A, the nonradiological hazardous materials were analyzed further because it was felt that the nonradiological hazardous material screening thresholds were too high.

Assessment of radiological and non-radiological hazards in the nuclear fuel cycle - The Indian experience

International Nuclear Information System (INIS)

Krishnamony, S.; Gopinath, D.V.

1996-01-01

Design and operational aspects of nuclear fuel cycle facilities have several features that distinguish them from nuclear power plants. These are related to (i) the nature of operations which are chiefly mining, metallurgical and chemical; (ii) the nature and type of radio-active materials handled, their specific activities and inventories; and (iii) the physical and chemical processes involved and the associated containment provisions. Generally the radioactive materials are present in an already highly dispersible or mobile form, in the form of solutions, slurries and powders, often associated with a wide variety of reactive and corrosive chemicals. There are further marked differences between the front-end and back-end of the fuel cycle. Whereas the front-end is characterized by the presence of large quantities of low specific activity naturally occurring radioactive materials, the back-end is characterized by high specific activities and concentrations of fission products and actinides. Radioactive characteristics of waste arisings are also different in different phases of the nuclear fuel cycle. Potential for internal exposure in the occupational environment is another distinguishing feature as compared with the more common designs of nuclear power reactors. Potential for accidents, their phenomenology and the resulting consequences are also markedly different in fuel cycle operations. The non-radiological hazards in fuel cycle operations are also of significance, since the operations are mostly mining, metallurgical and chemical in nature. These aspects are examined and evaluated in this paper, based on the Indian experience. (author). 12 refs, 10 tabs

Environmental and other evaluations of alternatives for long-term management of stored INEL transuranic waste

International Nuclear Information System (INIS)

1979-12-01

This study identifies, develops, and evaluates, in a preliminary manner, alternatives for long-term management of TRU waste stored at the Radioactive Waste Management Complex (RWMC) at the INEL. The evaluations concern waste currently at the RWMC and waste expected to be received by the beginning of the year 1985. The effects of waste that might be received after that data are addressed in an appendix. The technology required for managing the waste, the environmental effects, the risks to the public, the radiological and nonradiological hazards to workers, and the estimated costs are discussed

Environmental and other evaluations of alternatives for long-term management of stored INEL transuranic waste

Energy Technology Data Exchange (ETDEWEB)

1979-02-01

This study identifies, develops, and evaluates, in a preliminary manner, alternatives for long-term management of TRU waste stored at the Radioactive Waste Management Complex (RWMC) at the INEL. The evaluations concern waste currently at the RWMC and waste expected to be received by the beginning of the year 1985. The effects of waste that might be received after that date are addressed in an appendix. The technology required for managing the waste, the environmental effects, the risks to the public, the radiological and nonradiological hazards to workers, and the estimated costs are discussed.

Environmental and other evaluations of alternatives for long-term management of stored INEL transuranic waste

International Nuclear Information System (INIS)

1979-02-01

This study identifies, develops, and evaluates, in a preliminary manner, alternatives for long-term management of TRU waste stored at the Radioactive Waste Management Complex (RWMC) at the INEL. The evaluations concern waste currently at the RWMC and waste expected to be received by the beginning of the year 1985. The effects of waste that might be received after that date are addressed in an appendix. The technology required for managing the waste, the environmental effects, the risks to the public, the radiological and nonradiological hazards to workers, and the estimated costs are discussed

Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

Energy Technology Data Exchange (ETDEWEB)

Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Mayberry, J. [Science Applications International Corp., Idaho Falls, ID (United States); Frazier, G. [Univ. of Tennessee, Knoxville, TN (United States)

1994-01-01

This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well.

Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

International Nuclear Information System (INIS)

Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A.; Mayberry, J.; Frazier, G.

1994-01-01

This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well

ORNL necessary and sufficient standards for environment, safety, and health. Final report of the Identification Team for other industrial, radiological, and non-radiological hazard facilities

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-07-01

This Necessary and Sufficient (N and S) set of standards is for Other Industrial, Radiological, and Non-Radiological Hazard Facilities at Oak Ridge National Laboratory (ORNL). These facility classifications are based on a laboratory-wide approach to classify facilities by hazard category. An analysis of the hazards associated with the facilities at ORNL was conducted in 1993. To identify standards appropriate for these Other Industrial, Radiological, and Non-Radiological Hazard Facilities, the activities conducted in these facilities were assessed, and the hazards associated with the activities were identified. A preliminary hazards list was distributed to all ORNL organizations. The hazards identified in prior hazard analyses are contained in the list, and a category of other was provided in each general hazard area. A workshop to assist organizations in properly completing the list was held. Completed hazard screening lists were compiled for each ORNL division, and a master list was compiled for all Other Industrial, Radiological Hazard, and Non-Radiological facilities and activities. The master list was compared against the results of prior hazard analyses by research and development and environment, safety, and health personnel to ensure completeness. This list, which served as a basis for identifying applicable environment, safety, and health standards, appears in Appendix A.

ORNL necessary and sufficient standards for environment, safety, and health. Final report of the Identification Team for other industrial, radiological, and non-radiological hazard facilities

International Nuclear Information System (INIS)

1998-07-01

This Necessary and Sufficient (N and S) set of standards is for Other Industrial, Radiological, and Non-Radiological Hazard Facilities at Oak Ridge National Laboratory (ORNL). These facility classifications are based on a laboratory-wide approach to classify facilities by hazard category. An analysis of the hazards associated with the facilities at ORNL was conducted in 1993. To identify standards appropriate for these Other Industrial, Radiological, and Non-Radiological Hazard Facilities, the activities conducted in these facilities were assessed, and the hazards associated with the activities were identified. A preliminary hazards list was distributed to all ORNL organizations. The hazards identified in prior hazard analyses are contained in the list, and a category of other was provided in each general hazard area. A workshop to assist organizations in properly completing the list was held. Completed hazard screening lists were compiled for each ORNL division, and a master list was compiled for all Other Industrial, Radiological Hazard, and Non-Radiological facilities and activities. The master list was compared against the results of prior hazard analyses by research and development and environment, safety, and health personnel to ensure completeness. This list, which served as a basis for identifying applicable environment, safety, and health standards, appears in Appendix A

EU-CIS joint study project 2. Intervention criteria in CIS, risk assessments and non-radiological factors in decision-making

Energy Technology Data Exchange (ETDEWEB)

Hedemann Jensen, P. [Risoe National Lab., Roskilde (Denmark); Demin, V.F. [Russian Reserch Centre `Kurchatov Inst.`, Moscow (Russian Federation); Konstantinov, Y.O. [Research Inst. of Radiation Hygiene, St. Petersburg (Russian Federation); Likhtarev, I.A. [Ukrainian Scientific Centre for Radiation Medicine, Kiev (Ukraine); Rolevich, I.V. [Chernobyl State Commiettee, Minsk (Belarus); Schneider, T. [Centre d`etudes sur l`Evaluation de la Protection dans le domaine Nucleaire, CEPN, Paris (France)

1996-05-01

An extensive radiation risk estimation methodology has recently been developed in Russia and used for estimates of risk in exposed populations in the republics of Russia, Belarus and Ukraine. Results based on demographic data for the three republics are presented and compared with risk estimates from the EU risk model ASQRAD. The intervention criteria in the CIS republics have been evolving since the Chernobyl accident. The development of criteria in each of the three republics has been analysed and the CIS-Criteria have been compared to international guidance on intervention. After a nuclear or radiological emergency both radiological and non-radiological protection factors will influence the level of protective actions being introduced. The role of non-radiological protection factors in the overall optimization of health protection is addressed. It is argued that optimization of the overall health protection is not a question of developing radiation radiation protection philosophy to fully include socio-psychological factors. It is rather a question of including these factors - in parallel with the radiological protection factors - in cooperation between radiation protection experts and psychological specialists under the responsibility of the decision maker. (au) 19 tabs., 10 ills., 45 refs.

EU-CIS joint study project 2. Intervention criteria in CIS, risk assessments and non-radiological factors in decision-making

International Nuclear Information System (INIS)

Hedemann Jensen, P.; Demin, V.F.; Konstantinov, Y.O.; Likhtarev, I.A.; Rolevich, I.V.; Schneider, T.

1996-05-01

An extensive radiation risk estimation methodology has recently been developed in Russia and used for estimates of risk in exposed populations in the republics of Russia, Belarus and Ukraine. Results based on demographic data for the three republics are presented and compared with risk estimates from the EU risk model ASQRAD. The intervention criteria in the CIS republics have been evolving since the Chernobyl accident. The development of criteria in each of the three republics has been analysed and the CIS-Criteria have been compared to international guidance on intervention. After a nuclear or radiological emergency both radiological and non-radiological protection factors will influence the level of protective actions being introduced. The role of non-radiological protection factors in the overall optimization of health protection is addressed. It is argued that optimization of the overall health protection is not a question of developing radiation radiation protection philosophy to fully include socio-psychological factors. It is rather a question of including these factors - in parallel with the radiological protection factors - in cooperation between radiation protection experts and psychological specialists under the responsibility of the decision maker. (au) 19 tabs., 10 ills., 45 refs

Development of threshold guidance: National Low-Level Radioactive Waste Management Program

International Nuclear Information System (INIS)

1986-09-01

The current study has been conducted to provide DOE with a technical basis for the development of threshold guidance. The objective of the study was to develop the necessary background information and recommendations to assist the DOE in implementing the threshold limit concept for the disposal of DOE wastes at DOE facilities. The nature of low-level radioactive waste (LLW) varies greatly in both form and radionuclide content. While some low-level waste streams can contain substantial quantities of radioactive constituents, a potentially significant fraction of low-level waste is contaminated either very slightly or not at all. There is a strong likelihood that managing wastes with extremely low levels of radioactivity as nonradioactive waste would pose no significant safety problems and could result in substantial cost savings relative to its handling as LLW. Since all materials, including waste products, contain some radioactivity, it is necessary to distinguish between those wastes that would require disposal as LLW and those that have sufficiently low levels of radiological content to be managed according to their nonradiological properties. 131 refs., 9 figs., 24 tabs

Improving radiation awareness and feeling of personal security of non-radiological medical staff by implementing a traffic light system in computed tomography

Energy Technology Data Exchange (ETDEWEB)

Heilmaier, C.; Mayor, A.; Zuber, N.; Weishaupt, D. [Stadtspital Triemli, Zurich (Switzerland). Dept. of Radiology; Fodor, P. [Stadtspital Triemli, Zurich (Switzerland). Dept. of Anesthesiology and Intensive Care Medicine

2016-03-15

Non-radiological medical professionals often need to remain in the scanning room during computed tomography (CT) examinations to supervise patients in critical condition. Independent of protective devices, their position significantly influences the radiation dose they receive. The purpose of this study was to assess if a traffic light system indicating areas of different radiation exposure improves non-radiological medical staff's radiation awareness and feeling of personal security. Phantom measurements were performed to define areas of different dose rates and colored stickers were applied on the floor according to a traffic light system: green = lowest, orange = intermediate, and red = highest possible radiation exposure. Non-radiological medical professionals with different years of working experience evaluated the system using a structured questionnaire. Kruskal-Wallis and Spearman's correlation test were applied for statistical analysis. Fifty-six subjects (30 physicians, 26 nursing staff) took part in this prospective study. Overall rating of the system was very good, and almost all professionals tried to stand in the green stickers during the scan. The system significantly increased radiation awareness and feeling of personal protection particularly in staff with ? 5 years of working experience (p < 0.05). The majority of non-radiological medical professionals stated that staying in the green stickers and patient care would be compatible. Knowledge of radiation protection was poor in all groups, especially among entry-level employees (p < 0.05). A traffic light system in the CT scanning room indicating areas with lowest, in-termediate, and highest possible radiation exposure is much appreciated. It increases radiation awareness, improves the sense of personal radiation protection, and may support endeavors to lower occupational radiation exposure, although the best radiation protection always is to re-main outside the CT room during the scan.

Improving radiation awareness and feeling of personal security of non-radiological medical staff by implementing a traffic light system in computed tomography

International Nuclear Information System (INIS)

Heilmaier, C.; Mayor, A.; Zuber, N.; Weishaupt, D.; Fodor, P.

2016-01-01

Non-radiological medical professionals often need to remain in the scanning room during computed tomography (CT) examinations to supervise patients in critical condition. Independent of protective devices, their position significantly influences the radiation dose they receive. The purpose of this study was to assess if a traffic light system indicating areas of different radiation exposure improves non-radiological medical staff's radiation awareness and feeling of personal security. Phantom measurements were performed to define areas of different dose rates and colored stickers were applied on the floor according to a traffic light system: green = lowest, orange = intermediate, and red = highest possible radiation exposure. Non-radiological medical professionals with different years of working experience evaluated the system using a structured questionnaire. Kruskal-Wallis and Spearman's correlation test were applied for statistical analysis. Fifty-six subjects (30 physicians, 26 nursing staff) took part in this prospective study. Overall rating of the system was very good, and almost all professionals tried to stand in the green stickers during the scan. The system significantly increased radiation awareness and feeling of personal protection particularly in staff with ? 5 years of working experience (p < 0.05). The majority of non-radiological medical professionals stated that staying in the green stickers and patient care would be compatible. Knowledge of radiation protection was poor in all groups, especially among entry-level employees (p < 0.05). A traffic light system in the CT scanning room indicating areas with lowest, in-termediate, and highest possible radiation exposure is much appreciated. It increases radiation awareness, improves the sense of personal radiation protection, and may support endeavors to lower occupational radiation exposure, although the best radiation protection always is to re-main outside the CT room during the scan.

Preliminary analysis of the cost and risk of transporting nuclear waste to potential candidate commercial repository sites

International Nuclear Information System (INIS)

Wilmot, E.L.; Madsen, M.M.; Cashwell, J.W.; Joy, D.S.

1983-06-01

This report documents preliminary cost and risk analyses that were performed in support of the Nuclear Waste Terminal Storage (NWTS) program. The analyses compare the costs and hazards of transporting wastes to each of five regions that contain potential candidate nuclear waste repository sites being considered by the NWTS program. These regions are: the Gulf Interior Region, the Permian Basin, the Paradox Basin, Yucca Mountain, and Hanford. Two fuel-cycle scenarios were analyzed: once-through and reprocessing. Transportation was assumed to be either entirely by truck or entirely by rail for each of the scenarios. The results from the risk analyses include those attributable to nonradiological causes and those attributable to the radioactive character of the wastes being transported. 17 references

Aspects on the acceptance of waste for disposal in SFR

International Nuclear Information System (INIS)

Torstenfelt, Boerje

2006-01-01

When licensing a final repository for radioactive waste certain assumptions have to be made concerning the waste. These assumptions cover radionuclide inventory and nonradiological materials and its physical and chemical impact on the waste, the repository and on the environment. Development of new waste treatment systems and waste packages at the waste producer site aim at finding solutions and products that can be stored, transported and disposed of safely and are economically sound. This paper discusses some aspects concerning development of new or modified waste products. It highlights the importance of analysing the whole sequence in treatment, handling and disposing the waste. The process should be to find an optimal solution for the whole system, considering the fact that what is best in one step it not necessary best for the whole system, including the post closure issues. (author)

A Transparent Framework for guiding Radiological and Non-Radiological Contaminated Land Risk Assessments

International Nuclear Information System (INIS)

Lee, Alex; Mathers, Dan

2003-01-01

A framework is presented that may be used as a transparent guidance to both radiological and non-radiological risk assessments. This framework has been developed by BNFL, with external consultation, to provide a systematic approach for identifying key system drivers and to guide associated research packages in light of data deficiencies and sources of model uncertainty. The process presented represents an advance on existing working practices yet combines regulator philosophy to produce a robust, comprehensive, cost-effective and transparent work package. It aims at lending added confidence to risk models thereby adding value to the decision process

Cost effectiveness of below-threshold waste disposal at DOE sites

International Nuclear Information System (INIS)

Smith, C.F.; Cohen, J.J.

1987-01-01

A minimal health and environmental risk, limitations on disposal capacity, and the relatively high costs of low level waste (LLW) disposal are basic driving forces that lead to consideration of less restrictive disposal of wastes with very low levels of radiological contamination. The term threshold limit describes radioactive wastes that have sufficiently low-levels of radiological content to be managed according to their nonradiological properties. Given the efforts described elsewhere to provide guidance on the definition of below threshold (BT) doses and concentration levels, the purpose of this study was to quantify the resultant quantities, costs and cost effectiveness of BT disposal. For purposes of consistency with the previous demonstrations of the application of the threshold concept, available data for waste streams at the Idaho National Engineering Laboratory (INEL) and the Savannah River Plant (SRP) sites were collected and analyzed with regard to volumes, radionuclide concentrations, and disposal costs. From this information, quantities of BT waste, potential cost savings and cost effectiveness values were estimated. 1 reference, 5 tables

Characterization of secondary solid waste anticipated from the treatment of trench water from Waste Area Grouping 6 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

Kent, T.E.; Taylor, P.A.

1992-09-01

This project was undertaken to demonstrate that new liquid waste streams, generated as a consequence of closure activities at Waste Area Grouping (WAG) 6, can be treated adequately by existing wastewater treatment facilities at Oak Ridge National Laboratory (ORNL) without producing hazardous secondary solid wastes. Previous bench-scale treatable studies indicated that ORNL treatment operations will adequately remove the contaminants although additional study was required in order to characterize the secondary waste materials produced as a result of the treatment A 0.5-L/min pilot plant was designed and constructed to accurately simulate the treatment capabilities of ORNL fill-scale (490 L/min) treatment facilities-the Process Waste Treatment Plant (PWTP) and Nonradiological Wastewater Treatment Plant (NRWTP). This new test system was able to produce secondary wastes in the quantities necessary for US Environmental Protection Agency toxicity characteristic leaching procedure (TCLP) testing. The test system was operated for a 45-d test period with a minimum of problems and downtime. The pilot plant operating data verified that the WAG 6 trench waters can be treated at the PWTP and NRWTP to meet the discharge limits. The results of TCLP testing indicate that none of the secondary solid wastes will be considered hazardous as defined by the Resource Conservation and Recovery Act

Critique of rationale for transmutation of nuclear waste

International Nuclear Information System (INIS)

Smith, C.F.; Cohen, J.J.

1980-07-01

It has been suggested that nuclear transmutation could be used in the elimination or reduction of hazards from radioactive wastes. The rationale for this suggestion is the subject of this paper. The objectives of partitioning-transmutation are described. The benefits are evaluated. The author concludes that transmutation would appear at best to offer the opportunity of reducing an already low risk. This would not seem to be justifiable considering the cost. If non-radiological risks are considered, there is a negative total benefit

Cost effectiveness of below-threshold waste disposal at DOE sites

International Nuclear Information System (INIS)

Wickham, L.E.; Smith, C.F.; Cohen, J.J.

1986-01-01

Previous study has indicated the feasibility of establishing a threshold of concentration below which certain low-level (radioactive wastes) (LLW) could be safely handled and disposed of by conventional means such as landfills. Such below-threshold wastes have been synonymously termed de minimis or below regulatory concern (BRC) and can be deemed appropriate for management according to their nonradiological characteristics. The objective of this study was to determine the cost effectiveness for management and disposal of below-threshold waste at certain US Department of Energy sites. The sites selected for this study were the Idaho National Engineering Laboratory and Savannah River Laboratory. Cost-benefit analysis was used to determine the impacts, benefits, and potential cost advantages of establishing and implementing a threshold limit

Predisposal management of low and intermediate level radioactive waste. Safety guide

International Nuclear Information System (INIS)

2003-01-01

Radioactive waste is generated in the generation of electricity in nuclear power reactors 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 LLW. This Safety Guide deals with the safety issues associated with the predisposal management of LLW from nuclear fuel cycle facilities, large research and development installations and radioisotope production facilities. This includes all steps and activities in the management of waste, from its initial generation to its final acceptance at a waste disposal facility or the removal of regulatory control. The predisposal management of radioactive waste includes decommissioning. The term 'decommissioning' encompasses both the process of decommissioning a facility and the management of the waste that results (prior to its disposal). Recommendations on the process of decommissioning are provided in Refs. Recommendations on the management of the waste resulting from decommissioning are included in this Safety Guide. Although the mining and milling of uranium and thorium ores is part of the nuclear fuel cycle, the management of the operational waste (e.g. waste rock, tailings and effluent treatment waste) from these activities is not within the scope of this Safety Guide. The LLW that is

Oconee Nuclear Station, Units 1, 2, and 3. Annual operating report for 1976, volume 1: nonradiological environmental surveillance report; II: summary of operations

International Nuclear Information System (INIS)

1977-01-01

The non-radiological environmental surveillance program including thermal and chemical effluents, water quality, fish populations, benthos, fish impingement, gas-bubble disease, and plankton, fish larvae and fish egg entrainment is described. Information is also presented concerning operations, personnel radiation exposures, and fuel examinations

A preliminary analysis of the risk of transporting nuclear waste to potential candidate commercial repository sites

International Nuclear Information System (INIS)

Madsen, M.M.

1984-01-01

In accordance with the provisions of the Nuclear Waste Policy Act of 1982, environmental assessments for potential candidate sites are required to provide a basis for selection of the first site for disposal of commercial radioactive waste in deep geologic repositories. A preliminary analysis of the impacts of transportation for each of the five potential sites will be described. Transportation was assumed to be entirely by truck or entirely by rail in order to obtain bounding impacts. This paper presents both radiological and nonradiological risks for the once-through fuel cycle

Low-level radioactive waste disposal. Study of a conceptual nuclear energy center at Green River, Utah

International Nuclear Information System (INIS)

Card, D.H.; Hunter, P.H.; Barg, D.; de Souza, F.; Felthauser, K.; Winkler, V.; White, R.

1982-02-01

This document constitutes a segment of a feasibility study investigating the ramifications of constructing a nuclear energy center in an arid western region. In this phase of the study, the alternatives for disposing of the low-level waste on the site are compared with the alternative of transporting the waste to the nearest commercial waste disposal site for permanent disposal. Both radiological and nonradiological impacts on the local socioeconomic infrastructure and the environment are considered. Disposal on the site was found to cost considerably less than off-site disposal with only negligible impacts associated with the disposal option on either mankind or the environment

A literature-based preliminary characterization of risks in the nuclear waste management system

International Nuclear Information System (INIS)

Daling, P.M.; Rhoads, R.E.; Van Luik, A.E.

1990-04-01

The objectives of this study were to (1) review the literature containing information on risks in the nuclear waste management system and (2) use this information to develop preliminary estimates of the potential magnitudes of these risks. Information was collected on a broad range of risk categories to assist the US Department of Energy (DOE) in communicating information about the risks in the waste management system. The study, which was completed prior to passage of the Nuclear Waste Policy Amendments Act of 1987, examined all of the portions of the nuclear waste management system envisioned by the DOE in the 1985 ''Mission Plant for the Civilian Radioactive Waste Management Program.'' As such, there may be statements in this paper that are not consistent with current DOE positions. The scope of this paper includes the repository, the integral Monitored Retrievable Storage (MRS) facility, and the transportation system that supports the repository and the MRS facility. Based on the results of this analysis, it is concluded that the radiological risks in the waste management system are small relative to nonradiological risks and relative to the risks of exposure to natural background radiation. 6 refs., 2 figs., 2 tabs

A methodology for assessing social considerations in transport of low and intermediate level radioactive waste

International Nuclear Information System (INIS)

Allsop, R.E.; Banister, D.J.; Holden, D.J.; Bird, J.; Downe, H.E.

1986-05-01

A methodology is proposed for taking into account non-radiological social aspects of the transport of low and intermediate level radioactive waste when considering the location of disposal facilities and the transport of waste to such facilities from the sites where it arises. As part of a data acquisition programme, an attitudinal survey of a sample of people unconnected with any suggested site or transport route is proposed in order to estimate levels of concern felt by people of different kinds about waste transport. Probabilities of accident occurrence during transport by road and rail are also discussed, and the limited extent of quantified information about consequences of accidents is reviewed. The scope for malicious interference with consignments of waste in transit is considered. (author)

Environmental and other evaluations of alternatives for management of defense transuranic waste at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

1982-04-01

Appendices to this report contain the following information: INEL history of Waste Management; text of communications between Idaho and the federal government on long-term management; agency and public response to a proposed environmental impact statement; updated estimates on radiological releases from the slagging-pyrolysis incinerator; modeling studies of subsurface migration of radionuclides; nonradiological emissions and their environmental effects; methods for calculating radiological consequences; analysis of abnormal events in conceptual retrieval and processing operations; environmental contamination by accidental releases; hazards to waste management workers; environmental and other effects of rail and truck shipment of wastes; effects of hypothetical worst-case shipping accidents in urban areas; environmental and other effects of processing INEL transuranic waste at the offsite geological repository; and regulations applicable to INEL TRU waste management

Cost avoidance techniques through the Fernald controlled area trash segregation program and the RIMIA solid waste reduction program

International Nuclear Information System (INIS)

Menche, C.E.

1997-01-01

The Fernald Environmental Management Project is a Department of Energy owned facility that produced high quality uranium metals for military defense. The Fernald mission has changed from one of production to remediation. Remediation is intended to clean up legacy (primary) waste from past practices. Little opportunity is available to reduce the amount of primary waste. However, there is an opportunity to reduce secondary waste generation, primarily through segregation. Two programs which accomplish this are the Controlled Area Trash Segregation Program and the RIMIA Solid Waste Reduction Program. With these two programs now in place at the FEMP, it has been estimated that a 60% reduction has been achieved in unnecessary clean waste being disposed as Low Level Waste at the Nevada Test Site. The cost savings associated with these programs (currently 79,000 cubic feet, $428,000) could easily run into the millions of dollars based on the upcoming restoration activities to be undertaken. The segregation of non-radiological waste in the radiologically Controlled Area not only establishes a firm commitment to send only low-level radioactive waste to the Nevada Test Site, but also results in substantial cost avoidance

Characterization of Secondary Solid Wastes in Trench Water in Waste Area Grouping 6 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

Taylor, P.A.; Kent, T.E.

1994-02-01

This project was undertaken to demonstrate that new liquid waste streams, generated as a consequence of closure activities at Waste Area Grouping (WAG) 6 and other sites, can be treated at the existing wastewater treatment facilities at Oak Ridge National Laboratory (ORNL) to meet discharge requirements without producing hazardous secondary solid wastes. Previous bench and pilot-scale treatability studies have shown that ORNL treatment operations will adequately remove the contaminants and that the secondary solid wastes produced were not hazardous when treating water from two trenches in WAG 6. This study used WAG 6 trench water spiked with the minimum concentration of Toxicity Characteristics Leaching Procedure (TCLP) constituents (chemicals that can make a waste hazardous) found in any groundwater samples at ORNL. The Wastewater Treatment Test Facility (WTTF), a 0.5 L/min pilot plant that simulates the treatment capabilities of the Process Waste Treatment Plant (PWPT) and Nonradiological Wastewater Treatment Plant (NRWTP), was used for this test. This test system, which is able to produce secondary wastes in the quantities necessary for TCLP testing, was operated for a 59-d test period with a minimum of problems and downtime. The pilot plant operating data verified that WAG 6 trench waters, spiked with the minimum concentration of TCLP contaminants measured to date, can be treated at the PWTP and NRWTP to meet current discharge limits. The results of the TCLP analysis indicated that none of the secondary solid wastes produced during the treatment of these wastewaters will be considered hazardous as defined by the Resource Conservation and Recovery Act

Environmental surveillance for the INEL radioactive waste management complex. Annual report, 1979

International Nuclear Information System (INIS)

Wickham, L.E.; Janke, D.H.

1980-12-01

This document is the 1979 annual environmental surveillance report for the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory. Included are tabulated data from and discussions about routine radiological monitoring of atmospheric, hydrologic, geologic, and biotic environments of the RWMC. Also included are discussions of selected nonradiological pollutants (e.g., sodium, etc.). It is concluded that (a) RWMC operations have not adversely affected local, existing environments; (b) environmental conditions within the Transuranic Storage Area are not corrosive enough to adversely affect transuranic waste storage containers, and (c) the addition of lakebed soil to pit, trench, and soil test plot areas has altered the moisture cycle characteristic of RWMC soil

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

International Nuclear Information System (INIS)

1978-01-01

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

Environmental impact assessment and socio political issues of nuclear waste management

International Nuclear Information System (INIS)

Harmaajaervi, I.; Tolsa, H.

1997-09-01

The study is a part of the Publicly Administrated Nuclear Waste Management Research Programme (JYT2) which was carried out in 1994-1996. The principal goal of the research programme has been to provide the authorities with information and research results relevant for the safety of nuclear waste management in order to support the various activities of the authorities. The main emphasis of the research programme focuses on the disposal of spent fuel. In addition to nuclear waste research in the field of natural sciences and technology, the research program- me has focused mostly on societal issues associated with nuclear waste disposal facilities and on the non-radiological environmental effects in the environs of the disposal site. Some of the local effects are already revealed in the research phase, before any final decisions are made as to the selection of the disposal site. The study has focused primarily on local and regional issues. The statutory requirement to conduct environ- mental impact assessment (EIA) chiefly concerns those who are responsible for waste management, but the authorities also need to acquire systematic information in the field to support developing requirements for the content and scope of EIA procedure and preparedness to check the assessments made. This is a report of the first parts of the study in 1994-1995. The report deals with the subject matter generally based on earlier studies in Finland and other countries. The results of the study will be reported later

Waste Isolation Pilot Plant site environmental report for calendar year 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-06-01

US Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires each DOE facility that conducts significant environmental protection programs to prepare an Annual Site Environmental Report (ASER). The purpose of the ASER is to summarize environmental data in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts. This ASER not only documents the required data, it also documents new and continued monitoring and compliance activities during the 1994 calendar year. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP) (DOE/WIPP 94-024). The EMP defines a comprehensive set of parameters that must be monitored to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater, air, soil, and biotics are monitored for radiological and nonradiological activity levels. The baseline radiological surveillance program covers the broader geographic area that encompasses nearby ranches, villages, and cities. Nonradiological studies focus on the area immediately surrounding the WIPP site.

Waste Isolation Pilot Plant site environmental report for calendar year 1994

International Nuclear Information System (INIS)

1995-06-01

US Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires each DOE facility that conducts significant environmental protection programs to prepare an Annual Site Environmental Report (ASER). The purpose of the ASER is to summarize environmental data in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts. This ASER not only documents the required data, it also documents new and continued monitoring and compliance activities during the 1994 calendar year. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP) (DOE/WIPP 94-024). The EMP defines a comprehensive set of parameters that must be monitored to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater, air, soil, and biotics are monitored for radiological and nonradiological activity levels. The baseline radiological surveillance program covers the broader geographic area that encompasses nearby ranches, villages, and cities. Nonradiological studies focus on the area immediately surrounding the WIPP site

National Waste Terminal Storage repository in a bedded salt formation for spent unreprocessed fuel. Occupational exposure and health physics studies. KE report No. 78-21-R

International Nuclear Information System (INIS)

1978-09-01

This report includes the Occupational Exposure and Health Physics studies of the National Waste Terminal Storage Repository Number 2 (NWTSR2). Section 1 deals with occupational radiation exposures. The results of dose equivalent and dose commitment calculations are summarized. The man-rem total is a summation of all doses to all personnel throughout a year. It would have to be divided by the number of total personnel involved, to obtain an average annual dose per person. Section 2 presents the occupational exposures due to nonradiological pollutants. The activities of workers and the equipment used during the construction, storage/retrieval and decommissioning of the facility are outlined. Tabulations are presented of the substances (dusts, fumes, gases and vapors) and physical agents (heat, vibration, and nonionizing radiation) to which personnel will be exposed in various surface and underground work areas during construction and decommissioning, and during storage and retrieval operations. Some significant nonradiological occupational exposures are summarized. Section 3 outlines the health physics program for the NWTSR2 facility. It is important to initiate the health physics program one or two years before the facility is placed in operation to establish the radiation background levels at the site and its surrounding area, and to collect environmental samples, both on-site and off-site, prior to waste storage and retrieval operations. The health physics organization consists of 15 persons, including five health physicists

Waste Isolation Pilot Plant site environmental report for calendar year 1990

International Nuclear Information System (INIS)

1990-01-01

The US Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Operational Environmental Monitoring Plan (OEMP) monitors a comprehensive set of parameters in order to detect any potential environmental impacts and establish baselines for future quantitative environmental impact evaluations. Surface water and groundwater, soil, and biotics are measured for background radiation. Nonradiological environmental monitoring activities include meteorological, air quality, soil properties, and the status of the local biological community. Ecological studies focus on the immediate area surrounding the site with emphasis on the salt storage pile, whereas baseline radiological surveillance covers a broader geographic area including nearby ranches, villages, and cities. Since the WIPP is still in a preoperational state, no waste has been received; therefore, certain elements required by Order DOE 5400.1 are not presented in this report. 15 figs. 19 tabs

Waste Isolation Pilot Plant site environmental report for calendar year 1990

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The US Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Operational Environmental Monitoring Plan (OEMP) monitors a comprehensive set of parameters in order to detect any potential environmental impacts and establish baselines for future quantitative environmental impact evaluations. Surface water and groundwater, soil, and biotics are measured for background radiation. Nonradiological environmental monitoring activities include meteorological, air quality, soil properties, and the status of the local biological community. Ecological studies focus on the immediate area surrounding the site with emphasis on the salt storage pile, whereas baseline radiological surveillance covers a broader geographic area including nearby ranches, villages, and cities. Since the WIPP is still in a preoperational state, no waste has been received; therefore, certain elements required by Order DOE 5400.1 are not presented in this report. 15 figs. 19 tabs.

Approach to defining de minimis, intermediate, and other classes of radioactive waste

International Nuclear Information System (INIS)

Cohen, J.J.; Smith, C.F.

1986-01-01

This study has developed a framework within which the complete spectrum of radioactive wastes can be defined. An approach has been developed that reflects both concerns in the framework of a radioactive waste classification system. In this approach, the class of any radioactive waste stream is dependent on its degree of radioactivity and its persistence. To be consistent with conventional systems, four waste classes are defined. In increasing order of concern due to radioactivity and/or duration, these are: 1. De Minimis Wastes: This waste has such a low content of radioactive material that it can be considered essentially nonradioactive and managed according to its nonradiological characteristics. 2. Low-Level Waste (LLW): Maximum concentrations for wastes considered to be in this class are prescribed in 10CFR61 as wastes that can be disposed of by shallow land burial methods. 3. Intermediate Level Waste (ILW): This category defines a class of waste whose content exceeds class C (10CFR61) levels, yet does not pose a sufficient hazard to justify management as a high-level waste (i.e., permanent isolation by deep geologic disposal). 4. High-Level Waste: HLW poses the most serious management problem and requires the most restrictive disposal methods. It is defined in NWPA as waste derived from the reprocessing of nuclear fuel and/or as highly radioactive wastes that require permanent isolation

Environmental impact assessment of the Swedish high-level radioactive waste disposal system - examples of likely considerations

International Nuclear Information System (INIS)

1994-01-01

Sweden is investigating the feasibility of establishing a high-level radioactive waste (HLW) disposal system consisting of three components as follows: (1) Encapsulation facility, (2) system for transporting waste and (3) geologic repository. Swedish law requires that an Environmental Impact Assessment (EIA) be written for any planned action expected to have a significant impact on the environment. Before embarking on construction and operation of a HLW disposal system, the Swedish government will evaluate the expected environmental impacts to assure that the Swedish people and environmental will not be unduly affected by the disposal system. The EIA process requires that reasonable alternatives to the proposed action, including the 'zero' or 'no action' alternative, be considered so that the final approved plan for disposal will have undergone scrutiny and comparison of alternatives to arrive at a plan which is the best achievable given reasonable physical and monetary constraints. This report has been prepared by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for use by the Swedish Radiation Protection Institute (SSI). The purpose of this report is to establish a document which outlines the types of information which would be in an EIA for a three part disposal system like that envisioned by the Swedish Nuclear Fuel and Waste Management Company (SKB) for the disposal of Sweden's HLW. Technical information that would normally be included in an EIA is outlined in this document. The SSI's primary interest is in radiological impacts. However, for the sake of completeness and also to evaluate all environmental impacts in a single document, non-radiological impacts are also included. Swedish authorities other than the SSI may have interest in the non-radiological parts of the document. 26 refs

Recommendations to the NRC for review criteria for alternative methods of low-level radioactive waste disposal: Environmental monitoring and surveillance programs

International Nuclear Information System (INIS)

Denham, D.H.; Stenner, R.D.; Eddy, P.A.; Jaquish, R.E.; Ramsdell, J.V. Jr.

1988-07-01

Licensing of a facility for low-level radioactive waste disposal requires the review of the environmental monitoring and surveillance programs. A set of review criteria is recommended for the US Nuclear Regulatory Commission (NRC) staff to use in each monitoring phase---preoperational, operational, and post operational---for evaluating radiological and selected nonradiological parameters in proposed environmental monitoring and surveillance programs at low-level waste disposal facilities. Applicable regulations, industry standards, and technical guidance on low-level radioactive waste are noted throughout the document. In the preoperational phase, the applicant must demonstrate that the environmental monitoring program identifies radiation levels and radionuclide concentrations at the site and also provides adequate basic data on the disposal site. Data recording and statistical analyses for this phase are addressed

Waste Isolation Pilot Plant site environmental report, for calendar year 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

The U.S. Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires DOE facilities, that conduct environmental protection programs, to annually prepare a Site Environmental Report (SER). The purpose of the SER is to provide an abstract of environmental assessments conducted in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit. The content of this SER is not restricted to a synopsis of the required data, in addition, information pertaining to new and continued monitoring and compliance activities during the 1995 calendar year are also included. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP). The EMP provides inclusive guidelines implemented to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater. air, soil, and biotic matrices are monitored for an array of radiological and nonradiological factors. The baseline radiological surveillance program encompasses a broader geographic area that includes nearby ranches, villages, and cities. Most elements of nonradiological assessments are conducted within the geographic vicinity of the WIPP site.

Waste Isolation Pilot Plant site environmental report, for calendar year 1995

International Nuclear Information System (INIS)

1996-01-01

The U.S. Department of Energy (DOE) Order 5400.1 General Environmental Protection Program, requires DOE facilities, that conduct environmental protection programs, to annually prepare a Site Environmental Report (SER). The purpose of the SER is to provide an abstract of environmental assessments conducted in order to characterize site environmental management performance, to confirm compliance with environmental standards and requirements, and to highlight significant programs and efforts of environmental merit. The content of this SER is not restricted to a synopsis of the required data, in addition, information pertaining to new and continued monitoring and compliance activities during the 1995 calendar year are also included. Data contained in this report are derived from those monitoring programs directed by the Waste Isolation Pilot Plant (WIPP) Environmental Monitoring Plan (EMP). The EMP provides inclusive guidelines implemented to detect potential impacts to the environment and to establish baseline measurements for future environmental evaluations. Surface water, groundwater. air, soil, and biotic matrices are monitored for an array of radiological and nonradiological factors. The baseline radiological surveillance program encompasses a broader geographic area that includes nearby ranches, villages, and cities. Most elements of nonradiological assessments are conducted within the geographic vicinity of the WIPP site

Site environmental report for Calendar Year 1994 on radiological and nonradiological parameters

International Nuclear Information System (INIS)

1995-01-01

Battelle Memorial Institute's nuclear research facilities are currently being maintained in a surveillance and maintenance (S ampersand M) mode with continual decontamination and decommissioning (D ampersand D) activities being conducted under Department of Energy (DOE) Contract W-7405-ENG-92. These activities are referred to under the Contract as the Battelle Columbus Laboratories Decommissioning Project (BCLDP). Operations referenced in this report are performed in support of S ampersand M and D ampersand D activities. Battelle's King Avenue facility is not considered in this report to the extent that the West Jefferson facility is. The source term at the King Avenue site is a small fraction of the source term at the West Jefferson site. Off site levels of radionuclides that could be attributed to the west Jefferson and King Avenue nuclear operations wereindistinguishable from background levels at specific locations where air, water, and direct radiation measurements were performed. Environmental monitoring continued to demonstrate compliance by Battelle with federal, state and local regulations. Routine, nonradiological activities performed include monitoring liquid effluents and monitoring the ground water system for the West Jefferson North site. Samples of various environmental media including air, water, grass, fish, field and garden crops, sediment and soil were collected from the region surrounding the two sites and analyzed

Environmental impact assessment and socio political issues of nuclear waste management; Ydinjaetehuollon ympaeristoevaikutusten arviointi ja sosiopoliittiset kysymykset

Energy Technology Data Exchange (ETDEWEB)

Harmaajaervi, I; Tolsa, H [VTT Communities and Infrastructure, Espoo (Finland). Urban Planning; Vuori, S [VTT Energy, Espoo (Finland). Nuclear Energy; Litmanen, T [Jyvaeskylae Univ. (Finland)

1997-09-01

The study is a part of the Publicly Administrated Nuclear Waste Management Research Programme (JYT2) which was carried out in 1994-1996. The principal goal of the research programme has been to provide the authorities with information and research results relevant for the safety of nuclear waste management in order to support the various activities of the authorities. The main emphasis of the research programme focuses on the disposal of spent fuel. In addition to nuclear waste research in the field of natural sciences and technology, the research program- me has focused mostly on societal issues associated with nuclear waste disposal facilities and on the non-radiological environmental effects in the environs of the disposal site. Some of the local effects are already revealed in the research phase, before any final decisions are made as to the selection of the disposal site. The study has focused primarily on local and regional issues. The statutory requirement to conduct environ- mental impact assessment (EIA) chiefly concerns those who are responsible for waste management, but the authorities also need to acquire systematic information in the field to support developing requirements for the content and scope of EIA procedure and preparedness to check the assessments made. This is a report of the first parts of the study in 1994-1995. The report deals with the subject matter generally based on earlier studies in Finland and other countries. The results of the study will be reported later. 101 refs.

The IAEA's activities in the field of radioactive waste management

International Nuclear Information System (INIS)

Semenov, B.A.

1984-01-01

The IAEA has been concerned with radioactive waste management since its inception. Its programme in this area was expanded in the mid 1970s as questions related to the management and disposal of radioactive wastes came into focus in conjunction with the further industrial development of nuclear power. The objectives of the Agency's waste management programme are to assist its Member States in the safe and effective management of wastes by organizing the exchange and dissemination of information, providing guidance and technical assistance and supporting research. The current programme addresses all aspects of the industrial use of nuclear power under the aspects (a) technology of handling and treatment of wastes, (b) underground disposal of wastes, (c) environmental aspects of nuclear energy, including sea disposal of radioactive wastes. Systematic reviews have been made and publications issued concerning the technology of handling, treating, conditioning, and storing various categories of wastes, including liquid and gaseous wastes, wastes from nuclear power plants, spent fuel reprocessing and mining and milling of uranium ores, as well as wastes from decommissioning of nuclear facilities. As waste disposal is the current issue of highest interest, an Agency programme was set up in 1977 to develop a set of guidelines on the safe underground disposal of low-, intermediate- and high-level wastes in shallow ground, rock cavities or deep geological repositories. This programme will continue until 1990. Eleven Safety Series and Technical Documents and Reports have been published under this programme so far, which also addresses safety and other criteria for waste disposal. The environmental part of the waste management programme is concerned with the assessment of radiological and non-radiological consequences of discharges from nuclear facilities, including de minimis concepts in waste disposal and environmental models and data for radionuclide releases

Non-radiological factors and decision making on the radiological protection of the environment

International Nuclear Information System (INIS)

Simcock, A.

2002-01-01

'Non-radiological factors' can cover both physical and non-physical issues. As far as physical issues are concerned, the appropriate course is not to forget that radioactive substances have to be considered in the same way as other substances in respect of their non-radioactive properties. 'Damage to amenities' and 'interference with legitimate uses of the sea' are long-standing descriptions of the non-physical aspects of marine pollution and degradation. A framework for a taxonomy of the interests involved in such aspects is suggested, using the three dimensions of the degree of linkage to the marine environment, the nature of the interaction with the marine environment, and the economic nature of the interest concerned. Questions of remoteness also arise. A multi-dimensional analysis of the risks to the interests concerned is suggested. The dimension of 'public response' is particularly significant for the non-physical aspects of marine pollution and degradation. This dimension is complex, being influenced by 'fright factors' and subject to media amplification. These influences can include special local economic circumstances and past experiences. Finally, the process for integrating physical and non-physical factors is examined. Early consideration is recommended of how to achieve a transparent presentation of the issues and the way in which decisions are to be taken. (author)

Current status of the waste identification program at AECL's Chalk River Laboratories

International Nuclear Information System (INIS)

Csullog, G.W.; Edwards, N.W.; TerHuurne, M.A.

1998-01-01

The management of routine operating waste by Waste Management and Decommissioning (WM and D) at Atomic Energy of Canada Limited's (AECL) Chalk River Laboratories (CRL) is supported by the Waste Identification (WI) Program. The principal purpose of the WI Program is to minimize the cost and the effort associated with waste characterization and waste tracking, which are needed to optimize waste handling, storage and disposal. The major steps in the WI Program are: (1) identify and characterize the processes that generate the routine radioactive wastes accepted by WM and D - radioisotope production, radioisotope use, reactor operation, fuel fabrication, et cetera (2) identify and characterize the routine blocks of waste generated by each process or activity - the initial characterization is based on inference (process knowledge) (3) prepare customized, template data sheets for each routine waste block - templates contain information such as package type, waste material, waste type, solidifying agent, the average non-radiological contaminant inventory, the average radiological contaminant inventory, and the waste class (4) ensure generators 'use the right piece of paper with the right waste' when they transfer waste to WM and D - that is they use the correct template data sheets to transfer routine wastes, by: identifying and marking waste collection points in the generator's facility; ensuring that generators implement effective waste collection/segregation procedures; implementing standard procedures to transfer waste to WM and D; and, auditing waste collection and segregation within a generator's facility (5) determine any additional waste block characterization requirements (is anything needed beyond the original characterization by process knowledge?) This paper describes the WI Program, it provides an example of its implementation, and it summarizes the current status of its implementation for both CRL and non-CRL waste generators. (author)

Westinghouse Hanford Company effluent discharges and solid waste management report for calendar year 1989: 200/600 Areas

International Nuclear Information System (INIS)

Brown, M.J.; P'Pool, R.K.; Thomas, S.P.

1990-05-01

This report presents calendar year 1989 radiological and nonradiological effluent discharge data from facilities in the 200 Areas and the 600 Area of the Hanford Site. Both summary and detailed effluent data are presented. In addition, radioactive and nonradioactive solid waste storage and disposal data for calendar year 1989 are furnished. Where appropriate, comparisons to previous years are made. The intent of the report is to demonstrate compliance of Westinghouse Hanford Company-operated facilities with administrative control values for radioactive constituents and applicable guidelines and standards (including Federal permit limits) for nonradioactive constituents. 11 refs., 20 tabs

Preliminary characterization of risks in the nuclear waste management system based on information in the literature

International Nuclear Information System (INIS)

Daling, P.M.; Rhoads, R.E.; Van Luick, A.E.; Fecht, B.A.; Nilson, S.A.; Sevigny, N.L.; Armstrong, G.R.; Hill, D.H.; Rowe, M.; Stern, E.

1992-01-01

This document presents preliminary information on the radiological and nonradiological risks in the nuclear waste management system. The objective of the study was to (1) review the literature containing information on risks in the nuclear waste management system and (2) use this information to develop preliminary estimates of the potential magnitude of these risks. Information was collected on a broad range of risk categories to assist the US Department of Energy (DOE) in communicating information about the risks in the waste management systems. The study examined all of the portions of the nuclear waste management system currently expected to be developed by the DOE. The scope of this document includes the potential repository, the integral MRS facility, and the transportation system that supports the potential repository and the MRS facility. Relevant literature was reviewed for several potential repository sites and geologic media. A wide range of ''risk categories'' are addressed in this report: (1) public and occupational risks from accidents that could release radiological materials, (2) public and occupational radiation exposure resulting from routine operations, (3) public and occupational risks from accidents involving hazards other than radioactive materials, and (4) public and occupational risks from exposure to nonradioactive hazardous materials during routine operations. The report is intended to provide a broad spectrum of risk-related information about the waste management system. This information is intended to be helpful for planning future studies

Treatability tests on water from a low-level waste burial ground

International Nuclear Information System (INIS)

Taylor, P.A.

1990-01-01

Lab-scale treatability tests on trench water from a low-level waste burial ground have shown that the water can be successfully treated by existing wastewater treatment plants at Oak Ridge National Laboratory. Water from the four most highly contaminated trenches that had been identified to date was used in the treatability tests. The softening and ion exchange processes used in the Process Wastewater Treatment Plant removed Sr-90 from the trench water, which was the only radionuclide present at above the discharge limits. The air stripping and activated carbon adsorption processes used in the Nonradiological Wastewater Treatment Plant removed volatile and semi-volatile organics, which were the main contaminants in the trench water, to below detection limits. 6 refs., 2 figs., 7 tabs

Safety assessment for the transportation of NECSA's LILW to the Vaalputs waste disposal facility

International Nuclear Information System (INIS)

Maphoto, K.P.; Raubenheimer, E.; Swart, H.

2008-01-01

The transport safety assessment was carried out with a view to assess the impact on the environment and the people living in it, from exposure to radioactivity during transportation of the radioactive materials. It provides estimates of radiological risks associated with the envisaged transport scenarios for the road transport mode. This is done by calculating the human health impact and radiological risk from transportation of LILW along the R563 route, N14 and eventually to the Vaalputs National Waste Disposal Facility. Various parameters are needed by the RADTRAN code in calculating the human health impact and risk. These include: numbers of population densities following the routes undertaken, number of stops made, and the speed at which the transport will be traversing at towards the final destination. The human health impact with regard to the dose to the public, LCF and risk associated with transportation of Necsa's LILW to the Vaalputs Waste Disposal Facility by road have been calculated using RADTRAN 5 code. The results for both accident and incident free scenarios have shown that the overall risks are insignificant and can be associated with any non-radiological transportation. (authors)

Field test results for radioactive waste drum characterization with Waste Inspection Tomography (WIT)

Energy Technology Data Exchange (ETDEWEB)

Bernardi, R.T. [Bio-Imaging Research, Inc., Lincolnshire, IL (United States)

1997-11-01

This paper summarizes the design, fabrication, factory testing, evaluation and demonstration of waste inspection tomography (WIT). WIT consists of a self-sufficient, mobile semi-trailer for Non-Destructive Evaluation and Non-Destructive Assay (NDE/NDA) characterization of nuclear waste drums using X-ray and gamma-ray tomographic techniques. The 23-month WIT Phase I initial test results include 2 MeV Digital Radiography (DR), Computed Tomography (CT), Anger camera imaging, Single Photon Emission Computed Tomography (SPECT), Gamma-Ray Spectroscopy, Collimated Gamma Scanning (CGS), and Active and Passive Computed Tomography (A&PCT) using a 1.4 mCi source of {sup 166}Ho. These techniques were initially demonstrated on a 55-gallon phantom drum with three simulated waste matrices of combustibles, heterogeneous metals, and cement using check sources of gamma active isotopes. Waste matrix identification, isotopic identification, and attenuation-corrected gamma activity determination were all demonstrated nondestructively and noninvasively. Preliminary field tests results with nuclear waste drums are summarized. WIT has inspected drums with 0 to 20 grams plutonium 239. The minimum measured was 0.131 gram plutonium 239 in cement. 8 figs.

Disposal of flow-level radioactive waste in Belgium: A safety analysis for inorganic chemotoxic elements

International Nuclear Information System (INIS)

Mallants, D.; Volckaert, G.; Marivoet, J.; Neerdael, B.

2000-01-01

Low-level radioactive waste often contains large quantities of inorganic chemical substances. Due attention should therefore be given to the safety implications of both the radiological and chemical substances in the waste. Our study develops the safety assessment methodology for surface disposal with emphasis on the potential effects of inorganic nonradiological elements on human health. Contamination of groundwater was considered as the major exposure pathway. The applied methodology first screens all elements on the basis of five criteria. Conservative screening calculations were used to screen out the elements that do not pose danger to humans, and to select those that could have a negative impact and thus require further analysis. The latter was done by first calculating the elemental mass fluxes out of the repository and into the aquifer followed by the calculation of groundwater concentrations. The results showed that on the basis of the screening calculations, 75% of all elements could be classified as non-hazardous. The detailed calculations showed that the majority of the remaining elements had groundwater concentrations below the drinking water or groundwater standards. The results further showed that for a few elements the maximum groundwater concentration was above the standard, but below the background concentrations. (author)

Defense-Waste-Processing Faclity, Savannah River Plant, Aiken, SC: Draft environmental impact statement

International Nuclear Information System (INIS)

1981-09-01

The purpose of this Environmental Impact Statement (EIS) is to provide environmental input into both the selection of an appropriate strategy for the permanent disposal of the high-level radioactive waste (HLW) currently stored at the Savannah River Plant (SRP) and the subsequent decision to construct and operate a Defense Waste Processing Facility (DWPF) at the SRP site. The SRP is a major US Department of Energy (DOE) installation for the production of nuclear materials for national defense. Approximately 83 x 10 3 m 3 (22 million gal) of HLW currently are stored in tanks at the SRP site. The proposed DWPF would process the liquid HLW generated by SRP operations into a stable form for ultimate disposal. This EIS assesses the effects of the proposed immobilization project on land use, air quality, water quality, ecological systems, health risk, cultural resources, endangered species, wetlands protection, resource depletion, and regional social and economic systems. The radiological and nonradiological risks of transporting the immobilized wastes are assessed. The environmental impacts of disposal alternatives have recently been evaluated in a previous EIS and are therefore only summarized in this EIS

Defense Waste Processing Facility: Savannah River Plant, Aiken, SC. Final environmental impact statement

International Nuclear Information System (INIS)

1982-02-01

The purpose of this Environmental Impact Statement (EIS) is to provide environmental input into both the selection of an appropriate strategy for the permanent disposal of the high-level radioactive waste (HLW) currently stored at the Savannah River Plant (SRP) and the subsequent decision to construct and operate a Defense Waste Processing Facility (DWPF) at the SRP site. The SRP is a major US Department of Envgy (DOE) installation for the production of nuclear materials for national defense. Approximately 83 x 10 3 m 3 (22 million gal) of HLW currently are stored in tanks at the SRP site. The proposed DWPF would process the liquid HLW generated by SRP operations into a stable form for ultimate disposal. This EIS assesses the effects of the proposed immobilization project on land use, air quality, water quality, ecological systems, health risk, cultural resources, endangered species, wetlands protection, resource depletion, and regional social and economic systems. The radiological and nonradiological risks of transporting the immobilized wastes are assessed. The environmental impacts of disposal alternatives have recently been evaluated in a previous EIS and are therefore only summarized in this EIS

Safety assessment for the transportation of NECSA's LILW to the Vaalputs waste disposal facility

Energy Technology Data Exchange (ETDEWEB)

Maphoto, K.P.; Raubenheimer, E.; Swart, H. [Nuclear Liabilities Management, NECSA, P O Box 582, Pretoria, 0001 (South Africa)

2008-07-01

The transport safety assessment was carried out with a view to assess the impact on the environment and the people living in it, from exposure to radioactivity during transportation of the radioactive materials. It provides estimates of radiological risks associated with the envisaged transport scenarios for the road transport mode. This is done by calculating the human health impact and radiological risk from transportation of LILW along the R563 route, N14 and eventually to the Vaalputs National Waste Disposal Facility. Various parameters are needed by the RADTRAN code in calculating the human health impact and risk. These include: numbers of population densities following the routes undertaken, number of stops made, and the speed at which the transport will be traversing at towards the final destination. The human health impact with regard to the dose to the public, LCF and risk associated with transportation of Necsa's LILW to the Vaalputs Waste Disposal Facility by road have been calculated using RADTRAN 5 code. The results for both accident and incident free scenarios have shown that the overall risks are insignificant and can be associated with any non-radiological transportation. (authors)

Annual site environmental monitoring report for the Waste Isolation Pilot Plant, Calendar year 1985

International Nuclear Information System (INIS)

Reith, C.; Prince, K.; Fischer, T.; Rodriguez, A.; Uhland, D.; Winstanley, D.

1986-04-01

This is the first Annual Site Environmental Monitoring Report for the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. The WIPP project is operated by the US Department of Energy (DOE) for the purpose of providing a research and development facility to demonstrate the safe disposal of radioactive wastes generated by the defense activities of the U.S. Government. The report provides a comprehensive description of environmental activities at WIPP during Calendar Year 1985, including: a description of the WIPP project and its mission; a description of the local environment, including demographics; a summary of environmental program information, including an update on the status of environmental permits and compliance activities; a presentation of the findings of the Radiological Baseline Program (RBP), which is a program to characterize radionuclide activities in the environment around the WIPP site; and a summary of findings of the Ecological Monitoring Program (EMP), which examines non-radiological impacts of WIPP construction on the surrounding ecosystem. The WIPP facility is under construction, and will not receive radioactive wastes before October 1988. Therefore, this report describes the status of preoperational (as opposed to operational) environmental activities. 29 refs., 17 figs., 22 tabs

Waste Isolation Pilot Plant annual site environmental report for calendar year 1992

Energy Technology Data Exchange (ETDEWEB)

1993-12-31

The US Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Operational Environmental Monitoring Plan (OEMP) defined a comprehensive set of parameters which are monitored to detect potential environmental impacts and establish baselines for future environmental evaluations. Surface water and groundwater, air, soil, and biotics are monitored for radioactivity levels. Nonradiological environmental monitoring activities include air, water quality, soil properties, meteorological measurements and determination of the status of the local biological community. Ecological studies focus on the immediate area surrounding the WIPP site with emphasis on the salt storage pile. The baseline radiological surveillance covers a broader geographic area including nearby ranches, villages, and cities. Since the WIPP is still in its preoperational phase (i.e., no waste has been received) certain operational requirements of DOE Orders 5400.1, 5400.5, and the Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance (DOE/EH-0173T) are not relevant. Therefore, this report does not discuss items such as radionuclide emissions and effluents and subsequent doses to the public.

Waste Isolation Pilot Plant annual site environmental report for calendar year 1992

International Nuclear Information System (INIS)

1993-01-01

The US Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Operational Environmental Monitoring Plan (OEMP) defined a comprehensive set of parameters which are monitored to detect potential environmental impacts and establish baselines for future environmental evaluations. Surface water and groundwater, air, soil, and biotics are monitored for radioactivity levels. Nonradiological environmental monitoring activities include air, water quality, soil properties, meteorological measurements and determination of the status of the local biological community. Ecological studies focus on the immediate area surrounding the WIPP site with emphasis on the salt storage pile. The baseline radiological surveillance covers a broader geographic area including nearby ranches, villages, and cities. Since the WIPP is still in its preoperational phase (i.e., no waste has been received) certain operational requirements of DOE Orders 5400.1, 5400.5, and the Environmental Regulatory Guide for Radiological Effluent Monitoring and Environmental Surveillance (DOE/EH-0173T) are not relevant. Therefore, this report does not discuss items such as radionuclide emissions and effluents and subsequent doses to the public

The International Conference on Radioactive Waste Management

International Nuclear Information System (INIS)

1983-01-01

The IAEA has been concerned with radioactive waste management since its inception. Its programme in this area was expanded in the mid 1970s as questions related to the management and disposal of radioactive wastes came into focus in conjunction with the further industrial development of nuclear power. The objectives of the Agency's wastes management programme are to assist its Member States in the safe and effective management of wastes by organizing the exchange and dissemination of information, providing guidance and technical assistance and supporting research. The current programme addresses all aspects of the industrial use of nuclear power under the aspects (a) technology of handling and treatment of wastes, (b) underground disposal of wastes, (c) environmental aspects of nuclear energy, including sea disposal of radioactive wastes. Systematic reviews have been made and publications issued concerning the technology of handling, treating, conditioning, and storing various categories of wastes, including liquid and gaseous wastes, wastes from nuclear power plants, spent fuel reprocessing and mining and milling of uranium ores, as well as wastes from decommissioning of nuclear facilities. As waste disposal is the current issue of highest interest, an Agency programme was set up in 1977 to develop a set of guidelines on the safe underground disposal of low-, intermediate- and high-level wastes in shallow ground, rock cavities or deep geological repositories. This programme will continue until 1990. Eleven Safety Series and Technical documents and reports have been published under this programme so far, which also addresses safety and other criteria for waste disposal. The environmental part of the waste management programme is concerned with the assessment of radiological and non-radiological consequences of discharges from nuclear facilities, including de minimis concepts in waste disposal and environmental models and data for radionuclide releases. The Agency

MWIP: Surrogate formulations for thermal treatment of low-level mixed waste

International Nuclear Information System (INIS)

Bostick, W.D.; Hoffmann, D.P.; Stevenson, R.J.; Richmond, A.A.; Bickford, D.F.

1994-01-01

The category of sludges, filter cakes, and other waste processing residuals represent the largest volume of low-level mixed (hazardous and radioactive) wastes within the US Department of Energy (DOE) complex. Treatment of these wastes to minimize the mobility of contaminants, and to eliminate the presence of free water, is required under the Federal Facility Compliance Act agreements between DOE and the Environmental Protection Agency. In the text, we summarize the currently available data for several of the high priority mixed-waste sludge inventories within DOE. Los Alamos National Laboratory TA-50 Sludge and Rocky Flats Plant By-Pass Sludge are transuranic (TRU)-contaminated sludges that were isolated with the use of silica-based filter aids. The Oak Ridge Y-12 Plant West End Treatment Facility Sludge is predominantly calcium carbonate and biomass. The Oak Ridge K-25 Site Pond Waste is a large-volume waste stream, containing clay, silt, and other debris in addition to precipitated metal hydroxides. We formulate ''simulants'' for the waste streams described above, using cerium oxide as a surrogate for the uranium or plutonium present in the authentic material. Use of nonradiological surrogates greatly simplifies material handling requirements for initial treatability studies. The use of synthetic mixtures for initial treatability testing will facilitate compositional variation for use in conjunction with statistical design experiments; this approach may help to identify any ''operating window'' limitations. The initial treatability testing demonstrations utilizing these ''simulants'' will be based upon vitrification, although the materials are also amenable to testing grout-based and other stabilization procedures. After the feasibility of treatment and the initial evaluation of treatment performance has been demonstrated, performance must be verified using authentic samples of the candidate waste stream

Results of field testing of radioactive waste forms using lysimeters

International Nuclear Information System (INIS)

McConnell, J.W., Jr.; Rogers, R.D.; Jastrow, J.D.; Wickliff, D.S.

1992-01-01

The Field Lysimeter Investigation: Low-Level Waste Data Base Development Program is obtaining informaiton on the performance of radioactive waste in a disposal environment. Waste forms fabricated using ion-exchange resins from EPICOR-II prefilters employed in the cleanup of the Three Mile Island (TMI) Nuclear Power Station are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. In this paper, radionuclide releases from waste forms in the first six years of sampling are presented and discussed. Application of lysimeter data to use in performance assessment models is presented. Initial results from use of data in a performance assessment model are discussed

Health and Environmental Protection Standards for Uranium and Thorium Mill Tailings (40 CFR Part 192)

Science.gov (United States)

This regulation sets standards for the protection of public health, safety, and the environment from radiological and non-radiological hazards from uranium and thorium ore processing and disposal of associated wastes.

H12: Examination of safety assessment aims, procedures and results from a wider perspective

International Nuclear Information System (INIS)

Neall, F.B; Smith, P.A.

2004-04-01

Safety assessment (SA) are a familiar tool for the evaluation of disposal concepts for radioactive waste. There is, however, often confusion in the wider community about the aims, methods and results used in SA. This report aims to present the H12 SA in a way that makes the assessment process clearer and the implications of the results more meaningful both to workers within the SA field and to a wider technical audience. The reasonableness of the assessment results, the quality of the models and databases and redundancy within the natural and engineered barrier system have been considered. A number of recent and somewhat older SAs that address a range of different waste types, host rocks and disposal concepts have been considered, and comparisons made to H12. A further aim is to put both doses and timescales in a more meaningful context. It has been necessary to: consider ways of demonstrating the meaningfulness of calculations that give results for many thousands of years in the future; provide a framework timescale as a context for SA results over long times; demonstrate the smallness of the risk associated with the doses by comparison with other radiological and non-radiological risks. The perception of risk, which is a critical issue for public acceptance of radioactive waste disposal and must be considered when seeking to present safety assessment results 'in perspective' to a wider audience, is also discussed. It is concluded that H12 is comparable in many ways to assessments carried out internationally. Some assumptions are somewhat arbitrary reflecting the generic stage of the Japanese programme, and are likely to become better founded in future exercises. Nevertheless, H12 provides a clear and well-founded message that it is feasible to site and construct a safe repository from HLW in Japan. (author)

Results from simulated contact-handled transuranic waste experiments at the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Molecke, M.A.; Sorensen, N.R.; Krumhansl, J.L.

1993-01-01

We conducted in situ experiments with nonradioactive, contact-handled transuranic (CH TRU) waste drums at the Waste Isolation Pilot Plant (WIPP) facility for about four years. We performed these tests in two rooms in rock salt, at WIPP, with drums surrounded by crushed salt or 70 wt % salt/30 wt % bentonite clay backfills, or partially submerged in a NaCl brine pool. Air and brine temperatures were maintained at ∼40C. These full-scale (210-L drum) experiments provided in situ data on: backfill material moisture-sorption and physical properties in the presence of brine; waste container corrosion adequacy; and, migration of chemical tracers (nonradioactive actinide and fission product simulants) in the near-field vicinity, all as a function of time. Individual drums, backfill, and brine samples were removed periodically for laboratory evaluations. Waste container testing in the presence of brine and brine-moistened backfill materials served as a severe overtest of long-term conditions that could be anticipated in an actual salt waste repository. We also obtained relevant operational-test emplacement and retrieval experience. All test results are intended to support both the acceptance of actual TRU wastes at the WIPP and performance assessment data needs. We provide an overview and technical data summary focusing on the WIPP CH TRU envirorunental overtests involving 174 waste drums in the presence of backfill materials and the brine pool, with posttest laboratory materials analyses of backfill sorbed-moisture content, CH TRU drum corrosion, tracer migration, and associated test observations

The Radiation Effect to Waste Glass that Resulting of Vitrification

International Nuclear Information System (INIS)

Herlan Martono; Aisyah

2002-01-01

The high level liquid waste (HLLW) is generated from the first step extraction of the nuclear fuel reprocessing. This waste was contain of few of actinide and many of fission product. The alpha radiation of actinide that contain on the HLLW cause the change the waste glass characteristic. The experiment was conducted by the doping, irradiation and heating of waste glass resulting from vitrification. The alpha radiation cause the change of composition that could be detected from change of waste glass density and mechanical strength. The increasing of alpha radiation dose cause the increasing change of density and mechanical strength, although the change of mechanical strength is not significant. Degree of change of waste glass density also depend on type of waste-glass and reach for saturated point at over of 5x10 24 alpha decay/m 3 . The gamma radiation of fission product that contain on the HLLW can increasing of waste glass temperature that cause the structure change, so devitrification was occur. The devitrification can the increasing of leaching rate. The cumulative of gamma dose rate was not cause the devitrification. (author)

The Research Results of Radioactive Waste Management Technology Center Year 1997/1998

International Nuclear Information System (INIS)

1998-12-01

The research results of Radioactive Waste Management Technology Center, National Atomic Energy Agency of Indonesia year 1997/1998 contain paper as form of research results on radioactive waste management related fields. There were included many aspects such as radioactive waste processing, storage, decontamination, decommissioning, safety and environmental aspects. There are 26 papers indexed individually (ID)

The Research Results of Radioactive Waste Management Technology Center Year 1996/1997

International Nuclear Information System (INIS)

Budiman, P.; Martono, H.; Las, T.; Lubis, E.; Mulyanto; Wisnubroto, D. S.; Sucipta

1997-12-01

The research results of Radioactive Waste Management Technology Center, National Atomic Energy Agency of Indonesia year 1996/1997 contain paper as form of research results on radioactive waste management related fields. There were included many aspects such as radioactive waste processing, storage, decontamination, decommissioning, safety and environmental aspects. There are 24 papers and 12 short communications indexed individually(ID)

Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

Energy Technology Data Exchange (ETDEWEB)

Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

1993-03-01

A total of 6.05 x 10{sup 7} gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10{sup 6} gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system`s discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system`s discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10{sup 3} gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10{sup 4} gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10{sup 3} ft{sup 3}) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10{sup 2} ft{sup 3} of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10{sup 8} gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10{sup 7} gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10{sup 3} ft{sup 3}) of solid sludge generated by the precipitation/clarification process were removed from the filter press room.

Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

Energy Technology Data Exchange (ETDEWEB)

Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

1993-03-01

A total of 6.05 x 10[sup 7] gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10[sup 6] gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system's discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system's discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10[sup 3] gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10[sup 4] gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10[sup 3] ft[sup 3]) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10[sup 2] ft[sup 3] of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10[sup 8] gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10[sup 7] gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10[sup 3] ft[sup 3]) of solid sludge generated by the precipitation/clarification process were removed from the filter

Liquid and Gaseous Waste Operations Department annual operating report, CY 1992

International Nuclear Information System (INIS)

Gillespie, M.A.; Maddox, J.J.; Scott, C.B.

1993-03-01

A total of 6.05 x 10 7 gal of liquid waste was decontaminated by the Process Waste Treatment Plant (PWTP) ion exchange system during CY 1992. This averaged to 115 gpm throughout the year. When necessary, a wastewater sidestream of 50--80 gpm was treated through the use of a natural zeolite treatment system. An additional 8.00 x 10 6 gal (average of 15 gpm throughout the year) were treated by the zeolite system. Therefore, the average total flow treated at the PWTP for CY 1992 was 130 gpm. In mid-June, the zeolite system was repiped to allow it the capability to treat the ion exchange system's discharge due to rising Cs problems in the wastewater. While being used to treat the ion exchange system's discharge, it cannot treat a sidestream of wastewater. During the year, the regeneration of the cation exchange resins resulted in the generation of 7.83 x 10 3 gal of liquid low-level waste (LLLW) concentrate and 1.15 x 10 4 gal of LLLW evaporator feed. The head-end softening process (precipitation/clarification) generated 604 drums (4.40 x 10 3 ft 3 ) of solid low-level waste sludge. The zeolite treatment system generated approximately 8.40 x 10 2 ft 3 of spent zeolite resin, which was turned over to the Solid Waste Operations Department for disposal. See Table 1 for a monthly summary of activities at the PWTP. Figures 1, 2, 3, and 4 show a comparison of operations at the PWTP in 1992 with previous years. Figure 5 shows a comparison of annual rainfall at Oak Ridge National Laboratory (ORNL) since 1987. A total of 1.55 x 10 8 gal of liquid waste (average of 294 gpm throughout the year) was treated at the Nonradiological Wastewater Treatment Plant (NRWTP). Of this amount, 1.40 x 10 7 gal were treated by the precipitation/clarification process for removal of heavy metals. Twenty-five boxes (1.60 x 10 3 ft 3 ) of solid sludge generated by the precipitation/clarification process were removed from the filter press room

Comprehensive characterization and hazard assessment of the DOE-Niagara Falls storage site

International Nuclear Information System (INIS)

Anderson, T.L.; Dettorre, J.F.; Jackson, D.R.; Ausmus, B.S.

1981-06-01

A comprehensive radioecological and nonradiological characterization and hazards assessment was conducted on DOE-Niagara Falls Storage Site. Pitchblende residues and other low-level nuclear waste have been stored on the site since 1944. The most highly radioactive residues were stored in four abandoned buildings, while other wastes were deposited in pits or piled on surface soils on the Site. Several ditches were constructed on the Site to facilitate drainage or excess precipitation. Results of the study will permit the US DOE to form an appropriate remedial action plan for the Site

Long-term management of the existing radioactive wastes and residues at the Niagara Falls Storage Site. Draft Environmental Impact Statement

International Nuclear Information System (INIS)

1984-08-01

The statement assesses and compares several alternatives for long-term management of the existing radioactive wastes and residues at the Niagara Falls Storage Site (NFSS), Lewiston, New York. The alternatives include: (1) no action (continued interim storage at NFSS within a diked and capped containment area), (2) long-term management at NFSS (improved containment, with or without modified form of the residues), (3) long-term management at other DOE sites (Hanford, Washington, or Oak Ridge, Tennessee), and (4) offsite management of the residues at Hanford or Oak Ridge and either leaving the wastes at NFSS or removing them for disposal in the ocean. In addition to alternatives analyzed in depth, several options are also considered, including: other modifications of residue form, modification of the basic conceptual designs, other containment design options, transportation routes, and transportation modes. The radiological health effects (primarily increased risk of cancer) associated with long-term management of the wastes and residues are expected to be smaller than the nonradiological risks of occupational and transportation-related injuries and deaths. During the action period, the risk is highest for workers if all wastes and residues are moved to Hanford. The risk is highest for the general public if the residues are moved to Hanford and the wastes are moved to the ocean. Dispersal of the slightly contaminated wastes in the ocean is not expected to result in any significant impacts on the ocean environment or pose any significant radiological risk to humans. For all alternatives, if controls ceased, there would be eventual dispersion of the radioactive materials to the environment. If it is assumed that all controls cease, predicted time for loss of covers over the buried materials ranges from several hundred years to more than two million years, depending on the use of the land surface

National Waste Repository Novi Han operational safety analysis report. Safety assessment methodology

International Nuclear Information System (INIS)

2003-01-01

The scope of the safety assessment (SA), presented includes: waste management functions (acceptance, conditioning, storage, disposal), inventory (current and expected in the future), hazards (radiological and non-radiological) and normal and accidental modes. The stages in the development of the SA are: criteria selection, information collection, safety analysis and safety assessment documentation. After the review the facilities functions and the national and international requirements, the criteria for safety level assessment are set. As a result from the 2nd stage actual parameters of the facility, necessary for safety analysis are obtained.The methodology is selected on the base of the comparability of the results with the results of previous safety assessments and existing standards and requirements. The procedure and requirements for scenarios selection are described. A radiological hazard categorisation of the facilities is presented. Qualitative hazards and operability analysis is applied. The resulting list of events are subjected to procedure for prioritization by method of 'criticality analysis', so the estimation of the risk is given for each event. The events that fall into category of risk on the boundary of acceptability or are unacceptable are subjected to the next steps of the analysis. As a result the lists with scenarios for PSA and possible design scenarios are established. PSA logical modeling and quantitative calculations of accident sequences are presented

Dossier: management of nuclear wastes. Research, results

International Nuclear Information System (INIS)

Anon.

2001-01-01

The researches carried out since many years on nuclear wastes have led to two main ways of management: the long-term conditioning of radio-elements and their advanced separation. The French atomic energy commission (CEA) has chosen to take up also the transmutation challenge, a way to transform long-living radioactive wastes into short-living radioactive wastes or stable compounds. The transmutation programs are based both on simulation and experiments with a huge international collaboration. This dossier presents in a digest way the research activity carried out on nuclear wastes processing and management at the CEA. (J.S.)

Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mccloy, John S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lepry, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rodriguez, Carmen P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Windisch, Charles F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westman, Matthew P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rieck, Bennett T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lang, Jesse B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Olszta, Matthew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, David A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2014-01-17

The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

Nuclear wastes future: results and forecasting

International Nuclear Information System (INIS)

2001-01-01

Since many years, the CEA is greatly involved in the research programs on the long-dated management of radioactive wastes. This document presents the CEA development in the following domains: the spent fuel processing, the high separation process, the environmental behavior of wastes packages, the glass performance and the apatites behavior in the new matrix. (A.L.B.)

Long-term management of wastes resulting from dismantling operations. Storing the very low-level activity wastes at Morvilliers

International Nuclear Information System (INIS)

Duret, F.; Dutzer, M.; Beranger, V.; Lecoq, P.

2003-01-01

Extension of dismantling operations in France in the years to come poses the question of availability of long-term waste facility. Large amount of such wastes will be produced after progressive shutdown of the 58 pressurized water reactors now in operation, not before 2010. However, France is already confronted with dismantling of 9 power reactors (6 of which of gas cooled graphite type), the first reprocessing plant at Marcoule, as well as, dismantling of other installations, for instance the CEA reactors or laboratories. The systems of processing the dismantling waste are not different from those used for wastes resulting from nuclear operations. For the high-level or long-term intermediate level activity disposal the debates must start by 2006, as based on the results of the research conducted according to different provisions of the December 30, 1991 law. These wastes represent however small amounts from the dismantling (around 2000 t for the 9 reactors at shutdown) and they will be stored until a decision will be made. A specific storing system should be implemented by 2008-2010 for the graphite wastes (around 23,000 t) which contain significant amount of long-lived radioelements, although their gross activity is low. But the most significant amount will come from low-level or intermediate-level of short lifetime or from wastes of very low activity. The first category is stored at Storage Center at Aube (CSA), its capacity being of 1,000,000 m 3 of drums. The total volume stored by the end of 2002 amounted 136,500 m 3 with an annual delivering of 12-15,000 m 3 at design rate of 30,000 m 3 /y. This center will be able to absorb the flux increase resulting from dismantling of the decommissioned nuclear installations (around 50,000 t from the dismantling of the 9 power reactor). The Center at Aube can be also adapted for storing wastes of large sizes as for instance the lid of the reactor vessel. According to the French regulation, the wastes produced within a

Results after ten years of field testing low-level radioactive waste forms using lysimeters

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.; Sanford, W.E.; Larsen, I.L.; Sullivan, T.M.

1995-01-01

The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. Ion-exchange resins from a commercial nuclear power station were solidified into waste forms using portland cement and vinyl esterstyrene. These waste forms are being tested to: (a) obtain information on performance of waste forms in typical disposal environments, (b) compare field results with bench leach studies, (c) develop a low-level waste data base for use in performance assessment source term calculations, and (d) apply the DUST computer code to compare predicted cumulative release to actual field data. The program, funded by the Nuclear Regulatory Commission (NRC), includes observed radionuclide releases from waste forms in field lysimeters. The purpose of this paper is to present the experimental results of two lysimeter arrays over 10 years of operation, and to compare those results to bench test results and to DUST code predicted releases. Further analysis of soil cores taken to define the observed upward migration of radionuclides in one lysimeter is also presented

E-waste collection in Italy: Results from an exploratory analysis.

Science.gov (United States)

Favot, Marinella; Grassetti, Luca

2017-09-01

This study looks at the performance of household electrical and electronic waste (WEEE) collection in 20 Italian regions from 2008 to 2015. The impact of several explicative variables on the results of e-waste collection is evaluated. The independent variables are socio-economic and demographic ones (age, gender, household size, education level, migration and income) along with technical-organisational variables (population density, presence of metropoles, macro regions, characteristics of the territory, percentage of household waste collected separately and number of e-waste collection points). The results show that the presence of collection points, the percentage of household waste collected separately and the percentage of females are positively correlated with the kg collected per inhabitant per year. For example, a variation of 1% of input (presence of collection points) corresponds to a 0.25% variation in the output (collection results) while 1% difference in the percentage of females in the population corresponds to a 7.549% difference in the collection rate. Population density, instead, is negatively correlated. It is interesting to note that there is a discrepancy between the Southern regions and the Centre regions (the former have an outcome 0.66 times lower than the latter) while the Northern regions perform similarly to the Centre ones. Moreover, the first year (2008) had a very low performance compared to the following years when the scheme constantly improved, mainly due to the additional collection points available. The Stochastic Frontier Model allows for the identification of the optimal production function among the 20 Italian regions. The best performing region is Tuscany (in the Centre), followed by Sardinia and Sicily (in the South). Copyright © 2017. Published by Elsevier Ltd.

Creep of ocean sediments resulting from the isolation of radioactive wastes

International Nuclear Information System (INIS)

Dawson, P.R.; Chavez, P.F.; Lipkin, J.; Silva, A.J.

1980-01-01

Predictive models for the creep of deep ocean sediments resulting from the disposal of radioactive wastes are presented and preliminary observations of a program for evaluation of creep constitutive equation parameters are discussed. The models are used to provide calculated response of sediments under waste disposal conditions

Simulation for Teaching and Assessment of Nodule Perception on Chest Radiography in Nonradiology Health Care Trainees.

Science.gov (United States)

Auffermann, William F; Henry, Travis S; Little, Brent P; Tigges, Stefan; Tridandapani, Srini

2015-11-01

Simulation has been used as an educational and assessment tool in several fields, generally involving training of physical skills. To date, simulation has found limited application in teaching and assessment of skills related to image perception and interpretation. The goal of this pilot study was to evaluate the feasibility of simulation as a tool for teaching and assessment of skills related to perception of nodules on chest radiography. This study received an exemption from the institutional review board. Subjects consisted of nonradiology health care trainees. Subjects underwent training and assessment of pulmonary nodule identification skills on chest radiographs at simulated radiology workstations. Subject performance was quantified by changes in area under the localization receiver operating characteristic curve. At the conclusion of the study, all subjects were given a questionnaire with five questions comparing learning at a simulated workstation with training using conventional materials. Statistical significance for questionnaire responses was tested using the Wilcoxon signed rank test. Subjects demonstrated statistically significant improvement in nodule identification after training at a simulated radiology workstation (change in area under the curve, 0.1079; P = .015). Subjects indicated that training on simulated radiology workstations was preferable to conventional training methods for all questions; P values for all questions were less than .01. Simulation may be a useful tool for teaching and assessment of skills related to medical image perception and interpretation. Further study is needed to determine which skills and trainee populations may be most amenable to training and assessment using simulation. Copyright © 2015 American College of Radiology. Published by Elsevier Inc. All rights reserved.

Principles and practices in managing the wastes resulting from decommissioning

International Nuclear Information System (INIS)

Vladescu, Gabriela; Oprescu, Theodor; Niculae, Ortenzia; Stan, Camelia

2004-01-01

. The chapter 2.4 presents a proposal for constituting a statistical basis for radioactive materials classification and the associated measuring procedures. The chapter 2.5 illustrates the principles by applying them to classification of solid. liquid and gaseous radioactive materials and their assignation to one of the categories: excluded, excepted, clean or radioactive. The results of this study can be applied in classifying the radioactive wastes produced in Romania in different nuclear activities such as Cernavoda NPP operation, nuclear research, industry, and medicine, decommissioning of different nuclear facilities, etc

Results of Sludge Mobilization Testing at Hanford High Level Waste (HLW) Tank

International Nuclear Information System (INIS)

STAEHR, T.W.

2001-01-01

Waste stored in the Tank 241-AZ-101 at the US DOE Hanford is scheduled as the initial feed for high-level waste vitrification. Tank 241-AZ-101 currently holds over 3,000,000 liters of waste made up of a settled sludge layer covered by a layer of liquid supernant. To retrieve the waste from the tank, it is necessary to mobilize and suspend the settled sludge so that the resulting slurry can be pumped from the tank for treatment and vitrification. Two 223.8-kilowatt mixer pumps have been installed in Tank 241-AZ-101 to mobilize the settled sludge layer of waste for retrieval. In May of 2000, the mixer pumps were subjected to a series of tests to determine (1) the extent to which the mixer pumps could mobilize the settle sludge layer of waste, (2) if the mixer pumps could function within operating parameters, and (3) if state-of-the-art monitoring equipment could effectively monitor and quantify the degree of sludge mobilization and suspension. This paper presents the major findings and results of the Tank 241-AZ-101 mixer pump tests, based on analysis of data and waste samples that were collected during the testing. Discussion of the results focuses on the effective cleaning radius achieved and the volume and concentration of sludge mobilized, with both one and two pumps operating in various configurations and speeds. The Tank 241-AZ-101 mixer pump tests were unique in that sludge mobilization parameters were measured using actual waste in an underground storage tank at the hanford Site. The methods and instruments that were used to measure waste mobilization parameters in Tank 241-AZ-101 can be used in other tanks. It can be concluded from the testing that the use of mixer pumps is an effective retrieval method for the mobilization of settled solids in Tank 241-AZ-101

Groundwater monitoring at the waste isolation pilot plant

International Nuclear Information System (INIS)

Kehrman, R.; Broberg, K.; Tatro, G.; Richardson, R.; Dasczcyszak, W.

1990-01-01

This paper discusses the Groundwater Monitoring Program (GMP) being conducted at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The Regulatory and Environmental Programs (REP) section of the Environment, Safety and Health department (ES ampersand H) is responsible for conducting environmental monitoring at the WIPP. Groundwater monitoring is one of the ongoing environmental activities currently taking place. The REP section includes water-quality sampling and water-level monitoring. The WIPP Project is a research and development facility designed to demonstrate the safe disposal of defense-generated TRU and mixed waste in a geologic repository. The Salado Formation of Permian age serves as the repository medium. The Salado Formation consists of bedded salt and associated evaporites. The formation is 602 meters thick at the site area; the top surface is located at a subsurface depth of 262 meters (10). The repository lies at a subsurface depth of 655 meters. Water-quality sampling for physical, chemical, and radiological parameters has been an ongoing activity at the WIPP site for the past six years, and will continue through the life of the project. Data collected from this program to date, has been used by Sandia National Laboratories for site characterization and performance assessment work. The data has also been used to establish a baseline of preoperational radiological and nonradiological groundwater quality. Once the facility begins receiving waste, this baseline will be used to determine if the WIPP facility influences or alters groundwater quality over time. The water quality of a well is determined while the well is continuously pumped. Serial samples of the pumped water are collected and tested for pH, Eh, temperature, specific gravity, specific conductivity, alkalinity, chlorides, divalent cations, ferrous iron, and total iron. 13 refs., 4 figs., 1 tab

Plasma/arc melter review for vitrification of mixed wastes: Results

Energy Technology Data Exchange (ETDEWEB)

Eddy, T.L.; Soelberg, N.R.; Raivo, B.D. [MeltTran, Inc., Idaho Falls, ID (United States)

1995-12-31

In October of 1994, the Idaho Waste Treatment Program (IWTP) sponsored a workshop to review the results of a plasma/arc melter system preliminary design for treating mixed waste. Attention focused on (1) the melter design, (2) the offgas system design, and (3) the overall system design. The inclusion of feed preparation and handling systems, as well as monitoring and control systems, were considered premature until decisions regarding the melter and offgas treatment were resolved. The evaluation was based on the constraints of the transuranic-contaminated mixed waste in the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering Laboratory (INEL). Major factors are the retention of the transuranics in the basaltic slag, maintenance in a radioactive environment, reliability of components to prevent any major problems, upsets, or safety concerns, and the collection, elimination, or reduction of hazardous materials for appropriate stabilization. Several modifications were recommended by the group at large, discussed by the subcommittees, and accepted as the preferred options by the design team. Though all questions were not answered, the preferred systems for mixed waste treatment were the arc melters with graphite electrode systems with appropriate cooling which reduced maintenance and the possibility of eruptions that have occurred with plasma torches. Arc melters can also result in the minimum footprint and shielding. The preferred offgas systems were the wet/dry systems, that essentially eliminate the formation of carcinogenic compounds so they do not have to be destroyed down stream. This system also puts all of the particulate matter into one stream, instead of two.

Thermal plasma treatment of cell-phone waste : preliminary result

Energy Technology Data Exchange (ETDEWEB)

Ruj, B. [Central Mechanical Engineering Research Inst., Durgapur (India). Thermal Engineering Group; Chang, J.S.; Li, O.L. [McMaster Univ., Hamilton, ON (Canada). Dept. of Engineering Physics; Pietsch, G. [RWTH Aachen Univ., Aachen (Germany)

2010-07-01

The cell phone is an indispensable service facilitator, however, the disposal and recycling of cell phones is a major problem. While the potential life span of a mobile phone, excluding batteries, is over 10 years, most of the users upgrade their phones approximately four times during this period. Cell phone waste is significantly more hazardous than many other municipal wastes as it contains thousands of components made of toxic chemicals and metals like lead, cadmium, chromium, mercury, polyvinyl chlorides (PVC), brominated flame retardants, beryllium, antimony and phthalates. Cell phones also use many expensive rare metals. Since cell phones are made up of plastics, metals, ceramics, and trace other substances, primitive recycling or disposal of cell phone waste to landfills and incinerators creates irreversible environmental damage by polluting water and soil, and contaminating air. In order to minimize releases into the environment and threat to human health, the disposal of cell phones needs to be managed in an environmentally friendly way. This paper discussed a safer method of reducing the generation of syngas and hydrocarbons and metal recovery through the treatment of cell phone wastes by a thermal plasma. The presentation discussed the experiment, with particular reference to sample preparation; experimental set-up; and results four samples with different experimental conditions. It was concluded that the plasma treatment of cell phone waste in reduced condition generates gaseous components such as hydrogen, carbon monoxide, and hydrocarbons which are combustible. Therefore, this system is an energy recovery system that contributes to resource conservation and reduction of climate change gases. 5 refs., 2 tabs., 2 figs.

Design assessment for Melton Valley liquid low-level waste collection and transfer system upgrade project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1994-10-01

This project is designed for collecting liquid low level waste (LLLW) from generating points inside the Radioisotope Engineering and Development Center (Buildings 7920 and 7930) facility and transferring this waste to the Collection Tank (F-1800) in the new Monitoring and Control Station (MCS) facility. The LLLW is transferred to the MCS in a new, underground, jacketed, stainless steel piping system. The LLLW will then be transferred from Tank F-1800 through a new, underground, jacketed, stainless steel piping system that connects the existing Bethel Valley LLLW Collection System and the Evaporator Facility Service Tanks. The interface for the two systems will be at the existing Interconnecting Pipe Line (ICPL) Valve Box adjacent to the Nonradiological Wastewater Treatment Plant. The project scope consists of the following systems: (1) Building 7920 LLLW Collection System; (2) Building 7930 LLLW Collection System; (3) LLLW Underground Transfer System to MCS; (4) MCS Building (including all equipment contained therein); (5) LLLW Underground Transfer System to ICPL Valve Box; and (6) Leak detection system for jacketed piping systems (3) and (5)

Waste Minimization Improvements Achieved Through Six Sigma Analysis Result In Significant Cost Savings

International Nuclear Information System (INIS)

Mousseau, Jeffrey D.; Jansen, John R.; Janke, David H.; Plowman, Catherine M.

2003-01-01

Improved waste minimization practices at the Department of Energy's (DOE) Idaho National Engineering and Environmental Laboratory (INEEL) are leading to a 15% reduction in the generation of hazardous and radioactive waste. Bechtel, BWXT Idaho, LLC (BBWI), the prime management and operations contractor at the INEEL, applied the Six Sigma improvement process to the INEEL Waste Minimization Program to review existing processes and define opportunities for improvement. Our Six Sigma analysis team: composed of an executive champion, process owner, a black belt and yellow belt, and technical and business team members used this statistical based process approach to analyze work processes and produced ten recommendations for improvement. Recommendations ranged from waste generator financial accountability for newly generated waste to enhanced employee recognition programs for waste minimization efforts. These improvements have now been implemented to reduce waste generation rates and are producing positive results

Results after nine years of field testing low-level radioactive waste forms using lysimeters

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.; Sanford, W.E.; Sullivan, T.M.

1995-01-01

The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. Ion-exchange resins from a nuclear power station were solidified into waste forms using Portland cement and vinyl ester-styrene. These waste forms are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. This paper reviews radionuclide releases from those waste forms in the first 9 years of sampling. Included is a discussion of the recently discovered upward migration of radionuclides. Also, lysimeter data are applied to a performance assessment source term model, and initial results are presented

Results Of The Extraction-Scrub-Strip Testing Using An Improved Solvent Formulation And Salt Waste Processing Facility Simulated Waste

International Nuclear Information System (INIS)

Peters, T.; Washington, A.; Fink, S.

2012-01-01

The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D Cs in an ESS test, using the baseline solvent formulation and the typical waste feed, is ∼15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under

1995 Site environmental report Sandia National Laboratories, Albuquerque, New Mexico

Energy Technology Data Exchange (ETDEWEB)

Shyr, L.J.; Duncan, D. [eds.; Sanchez, R.

1996-09-01

This 1995 report contains data from routine radiological and non-radiological environmental monitoring activities. Summaries of significant environmental compliance programs in progress, such as National Environmental Policy Act documentation, environmental permits, environmental restoration and various waste management programs at Sandia National Laboratories in Albuquerque, New Mexico, are included.

1995 Site environmental report Sandia National Laboratories, Albuquerque, New Mexico

International Nuclear Information System (INIS)

Shyr, L.J.; Duncan, D.; Sanchez, R.

1996-09-01

This 1995 report contains data from routine radiological and non-radiological environmental monitoring activities. Summaries of significant environmental compliance programs in progress, such as National Environmental Policy Act documentation, environmental permits, environmental restoration and various waste management programs at Sandia National Laboratories in Albuquerque, New Mexico, are included

Results of the CRCPD survey of 1984 low-level radioactive waste: progress to mid-September, 1986

International Nuclear Information System (INIS)

Devine, T.L.

1987-01-01

The survey of 1984 low-level radioactive waste by the Conference of Radiation Control Program Directors, Inc., is the second such survey. The previous survey was for waste generated during 1982. The CRCPD survey of 1984 LLRW requested information concerning the license, the effluents and other on-site managed wastes, details of exported waste type, the capacity for storing waste prior to shipment and its average utilization during 1984. Details of the exported waste included waste type, processing and packaging, NRC class, burial site or broker to which the waste was sent, and anticipated waste generation by year and by class through 1989. Shortcomings of the questionnaire and preliminary results are discussed. Based on the results of the two surveys of low-level radioactive waste conducted by the CRCPD, and the serious discrepancies which exist between data on waste shipped by generators and that on waste received by disposal sites, the following recommendation is made. That a single, national repository be established for all data on the generation and ultimate disposition of low-level radioactive waste. 1 figure, 1 table

Ceramic waste form qualification using results from witness tubes

International Nuclear Information System (INIS)

O'Holleran, T.P.; Johnson, S.G.; Bateman, K.J.

2002-01-01

A ceramic waste form has been developed to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The ceramic waste form is prepared in a hot isostatic press (HIP). The use of small, easily fabricated HIP capsules called witness tubes has been proposed as a practical way to obtain representative samples of ceramic waste form material for process monitoring, waste form qualification, and archiving. Witness tubes are filled with the same material used to fill the corresponding HIP can, and are HIPed along with the HIP can. Relevant physical, chemical, and performance (leach test) data are analyzed and compared. Differences between witness tube and HIP can materials are shown to be statistically insignificant, demonstrating that witness tubes do provide ceramic waste form material representative of the material in the corresponding HIP can.

Nonradiological chemical pathway analysis and identification of chemicals of concern for environmental monitoring at the Hanford Site

International Nuclear Information System (INIS)

Blanton, M.L.; Cooper, A.T.; Castleton, K.J.

1995-11-01

Pacific Northwest's Surface Environmental Surveillance Project (SESP) is an ongoing effort tot design, review, and conducted monitoring on and off the Hanford site. Chemicals of concern that were selected are listed. Using modeled exposure pathways, the offsite cancer incidence and hazard quotient were calculated and a retrospective pathway analysis performed to estimate what onsite concentrations would be required in the soil for each chemical of concern and other detected chemicals that would be required to obtain an estimated offsite human-health risk of 1.0E-06 cancer incidence or 1.0 hazard quotient. This analysis indicates that current nonradiological chemical contamination occurring on the site does not pose a significant offsite human-health risk; the highest cancer incidence to the offsite maximally exposed individual was from arsenic (1.76E-10); the highest hazard quotient was chromium(VI) (1.48E-04). The most sensitive pathways of exposure were surfacewater and aquatic food consumption. Combined total offsite excess cancer incidence was 2.09E-10 and estimated hazard quotient was 2.40E-04. Of the 17 identified chemicals of concern, the SESP does not currently (routinely) monitor arsenic, benzo(a)pyrene, bis(2- ethylhexyl)phthalate (BEHP), and chrysene. Only 3 of the chemicals of concern (arsenic, BEHP, chloroform) could actually occur in onsite soil at concern high enough to cause a 1.0E-06 excess cancer incidence or a 1.0 hazard index for a given offsite exposure pathway. During the retrospective analysis, 20 other chemicals were also evaluated; only vinyl chloride and thallium could reach targeted offsite risk values

Radioactive waste disposal by UKAEA establishments during 1980 and associated environmental monitoring results

International Nuclear Information System (INIS)

Flew, E.M.

1981-09-01

This report gives details of the amounts of solid and liquid radioactive waste disposed of by the principal establishments of the UKAEA during 1980. Waste arising at the UKAEA Nuclear Power Development Laboratories at Windscale and Springfields, which are both situated on British Nuclear Fuels Ltd. (BNFL)-sites, is disposed of by BNFL and included in their authorisations. Discharges to atmosphere of airborne radioactive waste are also included in the report. A summary of the results of the environmental monitoring programmes carried out in connection with the radioactive waste discharges is given. (author)

Radioactive waste disposal by UKAEA establishments during 1978 and associated environmental monitoring results

International Nuclear Information System (INIS)

Flew, E.M.

1979-05-01

This report gives details of the amounts of solid and liquid radioactive waste disposed of by the principal establishments of the UKAEA during 1978. Waste arising at the UKAEA Nuclear Power Development Laboratories at Windscale and Springfields, which are both situated on British Nuclear Fuels Ltd. (BNFL) sites, is disposed of by BNFL and included in their authorisations. Discharges to atmosphere of airborne radioactive waste are also included in the report. A summary of the results of the environmental monitoring programmes carried out in connection with the radioactive waste discharges is given. (author)

Field Lysimeter Investigations - test results: Low-Level Waste Data Base Development Program: Test results for fiscal years 1994-1995

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Rodgers, R.D.; Hilton, L.D.; Neilson, R.M. Jr.

1996-06-01

The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the U.S. Nuclear Regulatory Commission (NRC), is (1) studying the degradation effects in EPICOR-II organic ion-exchange resins caused by radiation, (2) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified EPICOR-II resins, (3) obtaining performance information on solidified EPICOR-II ion-exchange resins in a disposal environment, and (4) determining the condition of EPICOR-II liners. Results of the final 2 (10 total) years of data acquisition from operation of the field testing are presented and discussed. During the continuing field testing, both portland type I-II cement and Dow vinyl ester-styrene waste forms are being tested in lysimeter arrays located at Argonne National Laboratory-East in Illinois and at Oak Ridge National Laboratory. The experimental equipment is described and results of waste form characterization using tests recommended by the NRC's open-quotes Technical Position on Waste Formclose quotes are presented. The study is designed to provide continuous data on nuclide release and movement, as well as environmental conditions, over a 20-year period. At the end of the tenth year, the experiment was closed down. Examination of soil and waste forms is planned to be conducted next and will be reported later

Research results of sewage sludge and waste oil disposal by entrained bed gasification

Energy Technology Data Exchange (ETDEWEB)

Schingnitz, M.; Goehler, P.; Wenzel, W.; Seidel, W. (Noell-DBI Energie- und Entsorgungstechnik GmbH, Freiberg (Germany))

1992-01-01

Presents results of gasifying sewage sludge and waste oil with the GSP technology, developed by the Freiberg Fuel Institute (FRG). The GSP reactor was developed in 1976 for gasification of pulverized brown coal. An industrial reactor of this design operated for over 5 years with a total coal throughput of more than 300,000 t. The design of the gasification generator and the flowsheet of a 3 MW experimental pilot plant for waste gasification are presented. The PCB content in the gasification sludge is 6.14 mg/kg, in waste oil - 160 mg/kg. Gasification takes place at high temperatures of more than 1,400 C for complete destruction of toxic pollutants. Gasification results compare composition of raw gas produced by gasification of brown coal, sewage sludge and waste oil. A detailed list of content of pollutants (PCDD, PCDF, PAH, dioxin and furan) in the gasification gas, in process waters and in solid residue of the process water is provided. It is concluded that the GSP gasification process is suitable for safe disposal of waste with toxic content. 3 refs.

Waste Isolation Pilot Plant: Site environmental report for calendar year 1991

International Nuclear Information System (INIS)

1991-01-01

The US Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Operational Environmental Monitoring Plan (OEMP) monitors a comprehensive set of parameters in order to detect any potential environmental Impacts and establish baselines for future quantitative environmental Impact evaluations. Surface water and groundwater, air, soil, and biotics are measured for background radiation. Nonradiological environmental monitoring activities include air quality, water quality, soil properties, meteorological, and the status of the local biological community. Ecological studies focus on the immediate area surrounding the site with emphasis on the salt storage pile, whereas baseline radiological surveillance covers a broader geographic area Including nearby ranches, villages, and cities. Since the WIPP is still in a preoperational state, and no waste has been received; certain elements required by DOE Order 5400.1 are not presented In this report. The most significant addition to the 1991 report is the inclusion of the first four appendices, the Radiological Baseline Program (DOE/WIPP 92-037), the Salt Impact Studies (DOE/WIPP 92-038), the Disturbed Land Reclamation Techniques (DOE/WIPP 92-039), and the Background Water Characterization for the WIPP (DOE/WIPP 92-013). These appendices are independently published and available to interested parties by the DOE reference number. These summaries will not be published in future ASER'S. They will, however, be referenced as a basis for evaluating similar data collected during the Test and subsequent Operational phases of the WIPP

Waste conditioning for tank heel transfer. Preliminary data and results

International Nuclear Information System (INIS)

Ebadian, M.A.

1999-01-01

This report summarizes the research carried out at Florida International University's Hemispheric Center for Environmental Technology (FIU-HCET) for the fiscal year 1998 (FY98) under the Tank Focus Area (TFA) project ''Waste Conditioning for Tank Slurry Transfer.'' The objective of this project is to determine the effect of chemical and physical properties on the waste conditioning process and transfer. The focus of this research consisted in building a waste conditioning experimental facility to test different slurry simulants under different conditions, and analyzing their chemical and physical properties. This investigation would provide experimental data and analysis results that can make the tank waste conditioning process more efficient, improve the transfer system, and influence future modifications to the waste conditioning and transfer system. A waste conditioning experimental facility was built in order to test slurry simulants. The facility consists of a slurry vessel with several accessories for parameter control and sampling. The vessel also has a lid system with a shaft-mounted propeller connected to an air motor. In addition, a circulation system is connected to the slurry vessel for simulant cooling and heating. Experimental data collection and analysis of the chemical and physical properties of the tank slurry simulants has been emphasized. For this, one waste slurry simulant (Fernald) was developed, and another two simulants (SRS and Hanford) obtained from DOE sites were used. These simulants, composed of water, soluble metal salts, and insoluble solid particles, were used to represent the actual radioactive waste slurries from different DOE sites. The simulants' chemical and physical properties analyzed include density, viscosity, pH, settling rate, and volubility. These analyses were done to samples obtained from different experiments performed at room temperature but different mixing time and strength. The experimental results indicate that the

Radioactive waste disposal by UKAEA establishments during 1979 and associated environmental monitoring results

International Nuclear Information System (INIS)

Flew, E.M.

1980-07-01

This report gives details of the amounts of solid and liquid radioactive waste disposed of by the principal establishments of the UKAEA during 1979. Waste arising at the UKAEA Nuclear Power Development Laboratories at Windscale and Springfields, which are both situated on British Nuclear Fuels Ltd. (BNFL) sites, is disposed of by BNFL and included in their authorisations. Discharges to atmosphere of airborne radioactive waste are also included in the report. A summary of the results of the environmental monitoring programmes carried out in connection with the radioactive waste discharges is given. To facilitate an appreciation of the standard of safety achieved, the discharges are, where appropriate, shown as a percentage of those authorised. In the case of atmospheric discharges no quantitative limits are yet specified in the authorisations, but the results and estimates of discharges from stacks are compared with Derived Working Limits (DWL's) (i.e. a limit derived from the dose limits recommended by The International Commission on Radiological Protection in such a way that compliance with it implies virtual certainty of compliance with the relevant dose limits). Environmental monitoring results are also compared with appropriate DWL's. The principles underlying the control of the discharge of radioactive waste to the environment are summarised in an Appendix to the report. (author)

Management of radiological and non-radiological risks in a decommissioning project

International Nuclear Information System (INIS)

Deboodt, Pascal

2002-01-01

real commitment of each partner. We think that the ALARA approach is a very good way to provide an adapted language as well as such commitment. Without any doubt, this approach is mainly responsible for the good results we got. Thirdly, as far as the removal of asbestos at BR3 is concerned, it is obvious that the 'radiological approach' has brought some technical improvements to the 'non radiological' approach. Examples can be found in the use of the masks, of in the daily control for potential contamination. But, on the other end, the workers of the BR3 installation are now more aware of the potential existence of other sources of risks and of the rules, which have to be followed in such cases. Working into the nuclear field leads sometimes to a lack of awareness regarding 'industrial risks'. Some questions are still remaining as 'open questions'. Some of these have still been pointed out. How do we have to optimise such operations where more than one 'recognized' risk is involved? How did we cope up to now with such 'interactive' situations? How far do we have to optimise? What's the meaning of 'optimisation' in such cases? These are examples of questions we hope to deal with during the discussions with partners from the radiological and non-radiological fields

Results of a hospital waste survey in private hospitals in Fars province, Iran

International Nuclear Information System (INIS)

Askarian, Mehrdad; Vakili, Mahmood; Kabir, Gholamhosein

2004-01-01

Hospital waste is considered dangerous because it may possess pathogenic agents and can cause undesirable effects on human health and the environment. In Iran, neither rules have been compiled nor does exact information exist regarding hospital waste management. The survey presented in this article was carried out in all 15 private hospitals of Fars province (Iran) from the total numbers of 50 governmental and private hospitals located in this province, in order to determine the amount of different kinds of waste produced and the present situation of waste management. The results indicated that the waste generation rate is 4.45 kg/bed/day, which includes 1830 kg (71.44%) of domestic waste, 712 kg (27.8%) of infectious waste, and 19.6 kg (0.76%) of sharps. Segregation of the different types of waste is not carried out perfectly. Two (13.3%) of the hospitals use containers without lids for on-site transport of wastes. Nine (60%) of the hospitals are equipped with an incinerator and six of them (40%) have operational problems with the incinerators. In all hospitals municipal workers transport waste outside the hospital premises daily or at the most on alternative days. In the hospitals under study, there aren't any training courses about hospital waste management and the hazards associated with them. The training courses that are provided are either ineffective or unsuitable. Performing extensive studies all over the country, compiling and enacting rules, establishing standards and providing effective personnel training are the main challenges for the concerned authorities and specialists in this field

HARVESTING EMSP RESEARCH RESULTS FOR WASTE CLEANUP

International Nuclear Information System (INIS)

Guillen, Donna Post; Nielson, R. Bruce; Phillips, Ann Marie; Lebow, Scott

2003-01-01

The extent of environmental contamination created by the nuclear weapons legacy combined with expensive, ineffective waste cleanup strategies at many U.S. Department of Energy (DOE) sites prompted Congress to pass the FY96 Energy and Water Development Appropriations Act, which directed the DOE to: ''provide sufficient attention and resources to longer-term basic science research, which needs to be done to ultimately reduce cleanup costs'', ''develop a program that takes advantage of laboratory and university expertise, and'' ''seek new and innovative cleanup methods to replace current conventional approaches which are often costly and ineffective.'' In response, the DOE initiated the Environmental Management Science Program (EMSP)-a targeted, long-term research program intended to produce solutions to DOE's most pressing environmental problems. EMSP funds basic research to lower cleanup cost and reduce risk to workers, the public, and the environment; direct the nation's scientific infrastructure towards cleanup of contaminated waste sites; and bridge the gap between fundamental research and technology development activities. EMSP research projects are competitively awarded based on the project's scientific, merit coupled with relevance to addressing DOE site needs. This paper describes selected EMSP research projects with long, mid, and short-term deployment potential and discusses the impacts, focus, and results of the research. Results of EMSP research are intended to accelerate cleanup schedules, reduce cost or risk for current baselines, provide alternatives for contingency planning, or provide solutions to problems where no solutions exist

Hazards assessment for the Hazardous Waste Storage Facility

International Nuclear Information System (INIS)

Knudsen, J.K.; Calley, M.B.

1994-04-01

This report documents the hazards assessment for the Hazardous Waste Storage Facility (HWSF) located at the Idaho National Engineering Laboratory. The hazards assessment was performed to ensure that this facility complies with DOE and company requirements pertaining to emergency planning and preparedness for operational emergencies. The hazards assessment identifies and analyzes hazards that are significant enough to warrant consideration in a facility's operational emergency management program. The area surrounding HWSF, the buildings and structures at HWSF, and the processes used at HWSF are described in this report. All nonradiological hazardous materials at the HWSF were identified (radiological hazardous materials are not stored at HWSF) and screened against threshold quantities according to DOE Order 5500.3A guidance. Two of the identified hazardous materials exceeded their specified threshold quantity. This report discusses the potential release scenarios and consequences associated with an accidental release for each of the two identified hazardous materials, lead and mercury. Emergency considerations, such as emergency planning zones, emergency classes, protective actions, and emergency action levels, are also discussed based on the analysis of potential consequences. Evaluation of the potential consequences indicated that the highest emergency class for operational emergencies at the HWSF would be a Site Area Emergency

Site environmental report for 1994. Environmental report, January--December 1994

International Nuclear Information System (INIS)

1994-01-01

This document is the 1994 site environmental report for the Rocky Flats Environmental Technology Site for January thru December. Compliance programs, radiological and nonradiological monitoring, and significant issues and events are described. In addition, the methodology for radiation dose assessment and the Environmental Restoration, Waste Management, and Quality Assurance programs are discussed

Converting Simulated Sodium-bearing Waste into a Single Solid Waste Form by Evaporation: Laboratory- and Pilot-Scale Test Results on Recycling Evaporator Overheads

Energy Technology Data Exchange (ETDEWEB)

Griffith, D.; D. L. Griffith; R. J. Kirkham; L. G. Olson; S. J. Losinski

2004-01-01

Conversion of Idaho National Engineering and Environmental Laboratory radioactive sodium-bearing waste into a single solid waste form by evaporation was demonstrated in both flask-scale and pilot-scale agitated thin film evaporator tests. A sodium-bearing waste simulant was adjusted to represent an evaporator feed in which the acid from the distillate is concentrated, neutralized, and recycled back through the evaporator. The advantage to this flowsheet is that a single remote-handled transuranic waste form is produced in the evaporator bottoms without the generation of any low-level mixed secondary waste. However, use of a recycle flowsheet in sodium-bearing waste evaporation results in a 50% increase in remote-handled transuranic volume in comparison to a non-recycle flowsheet.

Perceived risks of radioactive waste transport through Oregon: Results of a statewide survey

International Nuclear Information System (INIS)

MacGregor, D.; Slovic, P.; Mason, R.G.; Detweiler, J.; Binney, S.E.; Dodd, B.

1994-01-01

Transportation of hazardous materials, and particularly radioactive wastes, on public highways has become an important risk management issue. The unfavorability of public attitudes regarding hazardous and nuclear waste signals the potential for strong public opposition to programs for transporting these materials. This paper presents the results of a survey conducted to assess public reactions to a long-term nuclear waste transport program planned to follow a route through a portion of rural Oregon. The survey assessed a number of key risk perception issues, including perceived health and safety risks of nuclear waste transport, relative risks of transport vs. storage at an existing site, trust in state officials, and satisfaction with life in communities along the transport route. The survey identified a number of attitudes and concerns that need to be understood and considered by those in charge of designing and implementing the waste-transportation program. 22 refs., 1 fig., 5 tabs

HARVESTING EMSP RESEARCH RESULTS FOR WASTE CLEANUP

Energy Technology Data Exchange (ETDEWEB)

Guillen, Donna Post; Nielson, R. Bruce; Phillips, Ann Marie; Lebow, Scott

2003-02-27

The extent of environmental contamination created by the nuclear weapons legacy combined with expensive, ineffective waste cleanup strategies at many U.S. Department of Energy (DOE) sites prompted Congress to pass the FY96 Energy and Water Development Appropriations Act, which directed the DOE to: ''provide sufficient attention and resources to longer-term basic science research, which needs to be done to ultimately reduce cleanup costs'', ''develop a program that takes advantage of laboratory and university expertise, and'' ''seek new and innovative cleanup methods to replace current conventional approaches which are often costly and ineffective.'' In response, the DOE initiated the Environmental Management Science Program (EMSP)-a targeted, long-term research program intended to produce solutions to DOE's most pressing environmental problems. EMSP funds basic research to lower cleanup cost and reduce risk to workers, the public, and the environment; direct the nation's scientific infrastructure towards cleanup of contaminated waste sites; and bridge the gap between fundamental research and technology development activities. EMSP research projects are competitively awarded based on the project's scientific, merit coupled with relevance to addressing DOE site needs. This paper describes selected EMSP research projects with long, mid, and short-term deployment potential and discusses the impacts, focus, and results of the research. Results of EMSP research are intended to accelerate cleanup schedules, reduce cost or risk for current baselines, provide alternatives for contingency planning, or provide solutions to problems where no solutions exist.

Using slaughterhouse waste in a biochemical-based biorefinery - results from pilot scale tests.

Science.gov (United States)

Schwede, Sebastian; Thorin, Eva; Lindmark, Johan; Klintenberg, Patrik; Jääskeläinen, Ari; Suhonen, Anssi; Laatikainen, Reino; Hakalehto, Elias

2017-05-01

A novel biorefinery concept was piloted using protein-rich slaughterhouse waste, chicken manure and straw as feedstocks. The basic idea was to provide a proof of concept for the production of platform chemicals and biofuels from organic waste materials at non-septic conditions. The desired biochemical routes were 2,3-butanediol and acetone-butanol fermentation. The results showed that hydrolysis resulted only in low amounts of easily degradable carbohydrates. However, amino acids released from the protein-rich slaughterhouse waste were utilized and fermented by the bacteria in the process. Product formation was directed towards acidogenic compounds rather than solventogenic products due to increasing pH-value affected by ammonia release during amino acid fermentation. Hence, the process was not effective for 2,3-butanediol production, whereas butyrate, propionate, γ-aminobutyrate and valerate were predominantly produced. This offered fast means for converting tedious protein-rich waste mixtures into utilizable chemical goods. Furthermore, the residual liquid from the bioreactor showed significantly higher biogas production potential than the corresponding substrates. The combination of the biorefinery approach to produce chemicals and biofuels with anaerobic digestion of the residues to recover energy in form of methane and nutrients that can be utilized for animal feed production could be a feasible concept for organic waste utilization.

Requirements for a long-term safety certification for chemotoxic substances stored in a final storage facility for high radioactive and heat-generating radioactive waste in rock salt formations

International Nuclear Information System (INIS)

Tholen, M.; Hippler, J.; Herzog, C.

2007-01-01

Within the scope of a project funded by the German Federal Ministry of Economics and Technology (Bundesministerium fuer Wirtschaft und Technologie, BMWi), a safety certification concept for a future permanent final storage for high radioactive and heat-generating radioactive waste (HAW disposal facility) in rock salt formations is being prepared. For a reference concept, compliance with safety requirements in regard to operational safety as well as radiological and non-radiological protection objectives related to long-term safety, including ground water protection, will be evaluated. This paper deals with the requirements for a long-term safety certification for the purpose of protecting ground water from chemotoxic substances. In particular, longterm safety certifications for the permanent disposal of radioactive waste in a HAW disposal facility in rock salt formations and for the dumping of hazardous waste in underground storage facilities in rock salt formations are first discussed, followed by an evaluation as to whether these methods can be applied to the long-term safety certification for chemotoxic substances. The authors find it advisable to apply the long-term safety certification for underground storage facilities to the long-term safety certification for chemotoxic substances stored in a HAW disposal facility in rock salt formations. In conclusion, a corresponding certification concept is introduced. (orig.)

Generation and management of solid waste resulting from tourist activities of the Porto de Galinhas - P

Directory of Open Access Journals (Sweden)

Jaqueline Guimarães Santos

2015-04-01

Full Text Available The significant solid waste generation, coupled with the lack of proper management of the waste generated, has been one of the issues of concern and conducting research on the part of scholars in the field. Tourism as an activity that positively impacts and negativity a given location, has emerged as an activity that can generate a lot of waste, especially in periods of high season, considering the increase of people moving to the tourist destinations. Accordingly, this study aims to analyze the generation and management of solid waste resulting from tourism in Porto de Galinhas, PE. We performed an exploratory, descriptive, qualitative study, conducted in the form of a case study in Porto de Galinhas, PE. The data collection was done interviews together social actors, as well as non-participant observation during data collection. The results showed that tourism activities in Porto de Galinhas result in a high amount of solid waste, and these are directed to inappropriate places. Although fate presents a combination of recyclable materials, RECYCLE, reuses this not a significant amount, given the proportion of waste generated.

Nuclear energy and the environment

International Nuclear Information System (INIS)

El-Hinnawi, E.E.

1980-01-01

Chapters are presented concerning the environmental impact of mining and milling of radioactive ores, upgrading processes, and fabrication of nuclear fuels; environmental impacts of nuclear power plants; non-radiological environmental implications of nuclear energy; radioactive releases from nuclear power plant accidents; environmental impact of reprocessing; nuclear waste disposal; fuel cycle; and the future of nuclear energy

DEWATERING TREATMENT SCALE-UP TESTING RESULTS OF HANFORD TANK WASTES

International Nuclear Information System (INIS)

TEDESCHI AR

2008-01-01

This report documents CH2M HILL Hanford Group Inc. (CH2M HILL) 2007 dryer testing results in Richland, WA at the AMEC Nuclear Ltd., GeoMelt Division (AMEC) Horn Rapids Test Site. It provides a discussion of scope and results to qualify the dryer system as a viable unit-operation in the continuing evaluation of the bulk vitrification process. A 10,000 liter (L) dryer/mixer was tested for supplemental treatment of Hanford tank low-activity wastes, drying and mixing a simulated non-radioactive salt solution with glass forming minerals. Testing validated the full scale equipment for producing dried product similar to smaller scale tests, and qualified the dryer system for a subsequent integrated dryer/vitrification test using the same simulant and glass formers. The dryer system is planned for installation at the Hanford tank farms to dry/mix radioactive waste for final treatment evaluation of the supplemental bulk vitrification process

Removing Phosphate from Hanford High-Phosphate Tank Wastes: FY 2010 Results

Energy Technology Data Exchange (ETDEWEB)

Lumetta, Gregg J.; Braley, Jenifer C.; Edwards, Matthew K.; Qafoku, Odeta; Felmy, Andrew R.; Carter, Jennifer C.; MacFarlan, Paul J.

2010-09-22

The U.S. Department of Energy (DOE) is responsible for environmental remediation at the Hanford Site in Washington State, a former nuclear weapons production site. Retrieving, processing, immobilizing, and disposing of the 2.2 × 105 m3 of radioactive wastes stored in the Hanford underground storage tanks dominates the overall environmental remediation effort at Hanford. The cornerstone of the tank waste remediation effort is the Hanford Tank Waste Treatment and Immobilization Plant (WTP). As currently designed, the capability of the WTP to treat and immobilize the Hanford tank wastes in the expected lifetime of the plant is questionable. For this reason, DOE has been pursuing supplemental treatment options for selected wastes. If implemented, these supplemental treatments will route certain waste components to processing and disposition pathways outside of WTP and thus will accelerate the overall Hanford tank waste remediation mission.

Derivation of activity limits for the disposal of radioactive waste in near surface disposal facilities

International Nuclear Information System (INIS)

2003-12-01

Radioactive waste must be managed safely, consistent with internationally agreed safety standards. The disposal method chosen for the waste should be commensurate with the hazard and longevity of the waste. Near surface disposal is an option used by many countries for the disposal of radioactive waste containing mainly short lived radionuclides and low concentrations of long lived radionuclides. The term 'near surface disposal' encompasses a wide range of design options, including disposal in engineered structures at or just below ground level, disposal in simple earthen trenches a few metres deep, disposal in engineered concrete vaults, and disposal in rock caverns several tens of metres below the surface. The use of a near surface disposal option requires design and operational measures to provide for the protection of human health and the environment, both during operation of the disposal facility and following its closure. To ensure the safety of both workers and the public (both in the short term and the long term), the operator is required to design a comprehensive waste management system for the safe operation and closure of a near surface disposal facility. Part of such a system is to establish criteria for accepting waste for disposal at the facility. The purpose of the criteria is to limit the consequences of events which could lead to radiation exposures and in addition, to prevent or limit hazards, which could arise from non-radiological causes. Waste acceptance criteria include limits on radionuclide content concentration in waste materials, and radionuclide amounts in packages and in the repository as a whole. They also include limits on quantity of free liquids, requirements for exclusion of chelating agents and pyrophoric materials, and specifications of the characteristics of the waste containers. Largely as a result of problems encountered at some disposal facilities operated in the past, in 1985 the IAEA published guidance on generic acceptance

Characterization of a low-level radioactive waste grout: Sampling and test results

International Nuclear Information System (INIS)

Martin, P.F.C.; Lokken, R.O.

1992-12-01

WHC manages and operates the grout treatment facility at Hanford as part of a DOE program to clean up wastes stored at federal nuclear production sites. PNL provides support to the grout disposal program through pilot-scale tests, performance assessments, and formulation verification activities. in 1988 and 1989, over one million gallons of a low-level radioactive liquid waste was processed through the facility to produce a grout waste that was then deposited in an underground vault. The liquid waste was phosphate/sulfate waste (PSW) generated in decontamination of the N Reactor. PNL sampled and tested the grout produced during the second half of the PSW campaign to support quality verification activities prior to grout vault closure. Samples of grout were obtained by inserting nested-tube samplers into the grout slurry in the vault. After the grout had cured, the inner tube of the sampler was removed and the grout samples extracted. Tests for compressive strength, sonic velocity, and leach testing were used to assess grout quality; results were compared to those from pilot-scale test grouts made with a simulated PSW. The grout produced during the second half of the PSW campaign exceeded compressive strength and leachability formulation criteria. The nested tube samplers were effective in collecting samples of grout although their use introduced greater variability into the compressive strength data

Waste Preparation and Transport Chemistry: Results of the FY 2001 Studies

Energy Technology Data Exchange (ETDEWEB)

Hunt, R.D.

2002-03-25

of researchers from AEA Technology, Florida International University (FIU), Fluor Hanford, Mississippi State University (MSU), Oak Ridge National Laboratory (ORNL), and Savannah River Technology Center (SRTC) to evaluate various aspects of the waste preparation and transport chemistry. The majority of this effort was focused on saltcake dissolution and saltwell pumping. The results of the AEA Technology, FIU, and MSU studies of saltcake dissolution and slurry transfers for Hanford are discussed in detail in a companion report prepared by T. D. Welch in 2001 (ORNIJTM-2001097). Staff members at Fluor Hanford have continued to conduct saltcake dissolution tests on actual tank waste (documented in reports prepared by D. L. Herting in 2000 and 2001). It should be noted that full-scale saltcake dissolution at Hanford is scheduled to begin in FY 2002. While the Hanford effort is focused on the transfer of waste from one tank to another, the objective of the SRTC study is the formation of aluminosilicates at elevated temperatures, which are present in the waste evaporator.

Contaminant distributions at typical U.S. uranium milling facilities and their effect on remedial action decisions

International Nuclear Information System (INIS)

Hamp, S.; Dotson, P.W.

1995-01-01

Past operations at uranium processing sites throughout the US have resulted in local contamination of soils and ground water by radionuclides, toxic metals, or both. Understanding the origin of contamination and how the constituents are distributed is a basic element for planning remedial action decisions. This report describes the radiological and nonradiological species found in ground water at a typical US uranium milling facility. The report will provide the audience with an understanding of the vast spectrum of contaminants that must be controlled in planning solutions to the long-term management of these waste materials

Environmental report for rod storage. Volume V

International Nuclear Information System (INIS)

Danese, F.L.

1981-09-01

Volume V of this report examines the environmental impact of the rod consolidation program. The postulated, nonsite-specific, differential impacts are primarily additional occupational exposures due to the rod storage operations. Other potential radiological and nonradiological impacts that are identified and addressed are negligible. There are no increases in population exposures except those associated with transportation of spent fuel and waste material. The increased utilization of existing spent fuel storage space could result in a decrease in the nonrecoverable material resources lost to new permanent spent fuel storage

Health Physics and Medical Services report for 1986

International Nuclear Information System (INIS)

Burt, A.K.; Bird, R.W.

1987-09-01

A Health Physics and Medical Services report is presented for Harwell Laboratory for 1986. Health physics aspects covered include safety policy and organisation, monitoring results for the working environment and personnel, an analysis of radiological incidents and radioactive waste disposal, and protection of the public. Other non-radiological aspects of health and safety are briefly considered. The section on Medical Services contains details of the staffing, the types of medical examinations performed, the treatments received, work on the safety of asbestos and manmade mineral fibres and training and education programmes. (UK)

Methods and results of a probabilistic risk assessment for radioactive waste transports

International Nuclear Information System (INIS)

Lange, F.; Gruendler, D.; Schwarz, G.

1993-01-01

The radiological risk from accidents has been analyzed for the expected annual transport volume (3400 shipping units) of low and partially intermediate level radioactive wastes to be shipped to a final repository. In order to take account of these variable quantities and conditions a computer code was developed to simulate a wide spectrum of waste transport and accident configurations using Monte Carlo sampling techniques. Typically some 10.000 source terms were generated to represent possible releases of radionuclides from transport accidents. Accident events in which the integrity of waste packagings is retained and consequently no releases occur are included. Potential radiological consequences are then calculated for each of the release categories by using an accident consequence code which takes into account atmospheric dispersion statistics. Finally cumulative complementary frequency distributions of radiological consequences are generated by superposing the results for all release categories. Radiological consequences are primarily expressed as potential effective individual doses resulting from airborne and deposited radionuclides. The results of the risk analysis show that expected frequencies of effective doses comparable to the natural radiation exposure of one year are quite low and very low for potential radiation exposures in the range of 50 mSv. (J.P.N.)

Radiological protection from radioactive waste management in existing exposure situations resulting from a nuclear accident.

Science.gov (United States)

Sugiyama, Daisuke; Hattori, Takatoshi

2013-01-01

In environmental remediation after nuclear accidents, radioactive wastes have to be appropriately managed in existing exposure situations with contamination resulting from the emission of radionuclides by such accidents. In this paper, a framework of radiation protection from radioactive waste management in existing exposure situations for application to the practical and reasonable waste management in contaminated areas, referring to related ICRP recommendations was proposed. In the proposed concept, intermediate reference levels for waste management are adopted gradually according to the progress of the reduction in the existing ambient dose in the environment on the basis of the principles of justification and optimisation by taking into account the practicability of the management of radioactive waste and environmental remediation. It is essential to include the participation of relevant stakeholders living in existing exposure situations in the selection of reference levels for the existing ambient dose and waste management.

The influence of slaughterhouse waste on fermentative H2 production from food waste: Preliminary results

International Nuclear Information System (INIS)

Boni, Maria Rosaria; Sbaffoni, Silvia; Tuccinardi, Letizia

2013-01-01

Highlights: • Co-digestion process finalized to bio-H 2 production was tested in batch tests. • Slaughterhouse waste (SHW) and food waste (FW) were co-digested in different proportions. • The presence of SHW affected the H 2 production from FW. • When SHW ranging between 50% and 70% the H 2 production is improved. • SHW percentages above 70%, led to a depletion in H 2 production. - Abstract: The aim of this study was to evaluate the influence of slaughterhouse waste (SHW; essentially the skin, fats, and meat waste of pork, poultry, and beef) in a fermentative co-digestion process for H 2 production from pre-selected organic waste taken from a refectory (food waste [FW]). Batch tests under mesophilic conditions were conducted in stirred reactors filled with different proportions of FW and SHW. The addition of 60% and 70% SHW to a mixture of SHW and FW improved H 2 production compared to that in FW only, reaching H 2 -production yields of 145 and 109 ml gVS 0 -1 , respectively, which are 1.5–2 times higher than that obtained with FW alone. Although the SHW ensured a more stable fermentative process due to its high buffering capacity, a depletion of H 2 production occurred when SHW fraction was higher than 70%. Above this percentage, the formation of foam and aggregated material created non-homogenous conditions of digestion. Additionally, the increasing amount of SHW in the reactors may lead to an accumulation of long chain fatty acids (LCFAs), which are potentially toxic for anaerobic microorganisms and may inhibit the normal evolution of the fermentative process

New developments and improvements in processing of 'problematic' radioactive waste. Results of a coordinated research project 2003-2007

International Nuclear Information System (INIS)

2007-12-01

This report addresses a category of wastes termed 'problematic wastes', wastes for which safe, efficient and cost effective methods for processing are not readily available. Processing options for many of these are identified and addressed. Results presented, illustrate the strategy for breaking 'problematic' waste streams down into a sequence of 'standard' issues which are amenable to solution. Decision makers and facility managers faced with problematic waste streams should be able to use this information to identify and pursue solutions to meet their needs. In this report, processing options for a total of 27 problematic waste streams that were identified and addressed by the individual laboratories participating in the Coordinated Research Project are discussed. These waste streams covered an extremely broad spectrum, ranging from simple, one component aqueous solutions originating from a research laboratory to very complex aqueous concentrates of waste resulting from reprocessing activities or reactor operation. These challenging wastes included: waste contaminated by tritium, wastes containing transuranic elements, and solid health care waste. The range of aqueous wastes included those contaminated by organic complexing agents and surfactants to pure organic waste such as contaminated oil. Correspondingly, the scale of approaches and technologies used to address these wastes is very broad. Use of this report is likely to be most effective as an initial screening tool to identify technologies best able to meet specific waste management objectives in terms of the waste generated, the technical complexity, the available economic resources, the environmental impact considerations, and the desired end product (output) of the technology. The report should assist the user to compare technologies and to reach an informed decision based on safety, technological maturity, economics, and other local needs

Transport of radioactive wastes to the planned final waste repository Konrad: Radiation exposure resulting from normal transport and radiological risks from transport accidents

International Nuclear Information System (INIS)

Lange, F.; Fett, H.J.; Gruendler, D.; Schwarz, G.

1993-01-01

Radiation exposures of members of critical groups of the general population and of transport personnel resulting from normal transport of radioactive wastes to the planned final waste repository Konrad have been evaluated in detail. By applying probabilistic safety assessment techniques radiological risks from transport accidents have been analysed by quantifying potential radiation exposures and contaminations of the biosphere in connection with their expected frequencies of occurrence. The Konrad transport study concentrates on the local region of the waste repository, where all transports converge. (orig.) [de

Uncertainty and sensitivity results for pre-waste-emplacement groundwater travel time

International Nuclear Information System (INIS)

Kaplan, P.G.

1992-01-01

In this paper uncertainty and sensitivity analyses for pre-waste-emplacement groundwater travel time conducted. Although preliminary, a numbed of interesting results were obtained. Uncertainty in the ground water travel time statistics, as measured by the coefficient of variation, increases and then decrease as the modeled system transitions from matrix-dominated to fracture-dominated flow. The uncertainty analysis also suggests that the median, as opposed to the mean, may be a better indicator of performance with respect to the regulatory criterion. The sensitivity analysis shows a strong correlation between an effective fracture property, fracture porosity, and failure to meet the regulatory pre-waste-emplacement groundwater travel time criterion of 1,000 years

Knowledge, attitudes and behaviour regarding waste management options in Romania: results from a school questionnaire

Directory of Open Access Journals (Sweden)

Karin KOLBE

2014-12-01

The analysis revealed that knowledge is highly developed in Romania regarding the potential of recycling, while the concepts of waste management technologies are far less known about and understood. Landfill is seen as a problem for human health and the environment. However, recycling behaviour is low - partly as a result of limited possibilities. In general, the treatment hierarchy that is recommended in the "European waste hierarchy" is only partly reflected in students’ attitudes towards waste management options.

Results of field testing of waste forms using lysimeters

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Rogers, R.D.

1988-01-01

The purpose of the field testing task, using lysimeter arrays, is to expose samples of solidified resin waste to the actual physical, chemical, and microbiological conditions of disposal enviroment. Wastes used in the experiment include a mixture of synthetic organic ion exchange resins and a mixture of organic exchange resins and an inorganic zeolite. Solidification agents used to produce the 4.8-by 7.6-cm cylindrical waste forms used in the study were Portland Type I-II cement and Dow vinyl ester-styrene. Seven of these waste forms were stacked end-to-end and inserted into each lysimeter to provide a 1-L volume. There are 10 lysimeters, 5 at ORNL and 5 at ANL-E. Lysimeters used in this study were designed to be self-contained units which will be disposed at the termination of the 20-year study. Each is a 0.91-by 3.12-m right-circular cylinder divided into an upper compartment, which contains fill material, waste forms, and instrumentation, and an empty lower compartment, which collects leachate. Four lysimeters at each site are filled with soil, while a fifth (used as a control) is filled with inert silica oxide sand. Instrumentation within each lysimeter includes porous cup soil-water samplers and soil moisture/temperature probes. The probes are connected to an on-site data acquisition and storage system (DAS) which also collects data from a field meteorological station located at each site. 9 refs

The impact of retrievability on disposal of radioactive waste

International Nuclear Information System (INIS)

Gera, F.; Hill, M.

2000-01-01

There are discussions in various countries about whether and to what degree the ability to retrieve wastes might be built into geological repositories for long lived radioactive waste. It is generally accepted that repositories should be designed so that retrieval will never be necessary on safety grounds. Nevertheless, reasons for retrieval have been put forward. The ways in which retrievability might be built into geological repositories and various expected impacts of such actions are briefly discussed in this paper. A preliminary comparison of some notional geological disposal strategies with varying degrees of retrievability is proposed. The comparison is qualitative because at present there are few detailed designs for geological repositories with retrievability and few assessments of the safety and other aspects of such repositories. The comparison has the aims of highlighting those factors that differ most from one strategy to another and identifying which of these factors require further assessment in order to make more complete and quantitative comparisons. The framework used for the preliminary comparison is that of a multi-attribute analysis, such as might be employed in an environmental impact assessment (EIA). This type of framework is chosen because it would be used in many countries to aid decisions between disposal options or strategies. The framework encompasses radiological and nuclear safety factors but goes well beyond these and includes a number of factors that are not quantifiable in the technical sense. However, consideration of such factors is considered important because they can have a significant impact on decision-making. The groups of factors are: radiological, nuclear safety and financial (quantifiable factors); non-radiological environmental impacts (partially quantifiable factors); non-quantifiable attributes (such as ethical and societal factors). (author)

Results of two years' operation of the waste processing cell PROLIXE

International Nuclear Information System (INIS)

Lecomte, M.; Madic, C.; Broudic, J.C.

1990-01-01

Solid wastes, contaminated by alpha, beta, gamma radioisotopes, are produced by spent fuel reprocessing and isotope production. The PROLIXE plant, prototype for leaching and encapsulation was put into operation in March 1988 for waste management with the following aims: development of decontamination by oxidative leaching of alpha wastes, to obtain less than 0.1 Ci/t for surface storage; recycling radioactive isotope recovered especially transuranium elements; define a versatile process for various solid radioactive waste for an industrial plant [fr

Initial Laboratory-Scale Melter Test Results for Combined Fission Product Waste

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Crum, Jarrod V.; Buchmiller, William C.; Rieck, Bennett T.; Schweiger, Michael J.; Vienna, John D.

2009-10-01

This report describes the methods and results used to vitrify a baseline glass, CSLNTM-C-2.5 in support of the AFCI (Advanced Fuel Cycle Initiative) using a Quartz Crucible Scale Melter at the Pacific Northwest National Laboratory. Document number AFCI-WAST-PMO-MI-DV-2009-000184.

Savannah River Site environmental data for 1993

Energy Technology Data Exchange (ETDEWEB)

Arnett, M.W. [ed.

1994-05-01

The figures and tables in this report represent a capsule view of the routine environmental monitoring and surveillance programs at the Savannah River Site. An attempt has been made to include all available data from environmental research programs. The first section of the book is a collection of maps of radiological and nonradiological sampling locations. Also included are general radiological and nonradiological sampling and analysis schedules; a list of the media sampled, along with sample sizes and representative aliquots; a list of the lower limits of detection for radiological detection instruments; the minimum detectable concentrations for gamma analysis of water and air samples; and the minimum detectable concentrations for gamma analysis of soil, food, fish and wildlife, and vegetation samples. Following the first section are data tables containing radiological and nonradiological effluent monitoring results, radiological and nonradiological environmental surveillance results, dose estimates, quality assurance activities, and results of nonroutine occurrences and special surveys.

The waste-to-energy framework for integrated multi-waste utilization: Waste cooking oil, waste lubricating oil, and waste plastics

Energy Technology Data Exchange (ETDEWEB)

Singhabhandhu, Ampaitepin; Tezuka, Tetsuo [Energy Economics Laboratory, Department of Socio-Environmental Energy Science, Graduate School of Energy Science, Kyoto University, Yoshida-honmachi, Sakyo-ku, Kyoto 606-8501 (Japan)

2010-06-15

Energy generation by wastes is considered one method of waste management that has the benefit of energy recovery. From the waste-to-energy point of view, waste cooking oil, waste lubricating oil, and waste plastics have been considered good candidates for feedstocks for energy conversion due to their high heating values. Compared to the independent management of these three wastes, the idea of co-processing them in integration is expected to gain more benefit. The economies of scale and the synergy of co-processing these wastes results in higher quality and higher yield of the end products. In this study, we use cost-benefit analysis to evaluate the integrated management scenario of collecting the three wastes and converting them to energy. We report the total heat of combustion of pyrolytic oil at the maximum and minimum conversion rates, and conduct a sensitivity analysis in which the parameters of an increase of the electricity cost for operating the process and increase of the feedstock transportation cost are tested. We evaluate the effects of economy of scale in the case of integrated waste management. We compare four cases of waste-to-energy conversion with the business as usual (BAU) scenario, and our results show that the integrated co-processing of waste cooking oil, waste lubricating oil, and waste plastics is the most profitable from the viewpoints of energy yield and economics. (author)

Demonstration Results on the Effects of Mercury Speciation on the Stabilization of Wastes

International Nuclear Information System (INIS)

Conley, T.B.; Hulet, G.A.; Morris, M.I.; Osborne-Lee, I.W.

1999-01-01

Mercury-contaminated wastes are currently being stored at approximately 19 Department of Energy sites, the volume of which is estimated to be about 16m(sup)3. These wastes exist in various forms including soil, sludges, and debris, which present a particular challenge regarding possible mercury stabilization methods. This reports provides the test results of three vendors, Allied Technology Group, IT Corporation, and Nuclear Fuel Services, Inc., that demonstrate the effects of mercury speciation on the stabilization of the mercury wastes. Mercury present in concentrations that exceed 260 parts per million must be removed by extraction methods and requires stabilization to ensure that the final wasteforms leach less than 0.2mg/L of mercury by the Toxicity Characteristic Leaching Procedure or 0.025 mg/L using the Universal Treatment Standard

FY-87 packing fabrication techniques (commercial waste form) results

International Nuclear Information System (INIS)

Werry, E.V.; Gates, T.E.; Cabbage, K.S.; Eklund, J.D.

1988-04-01

This report covers the investigation of fabrication techniques associated with the development of suitable materials and methods to provide a prefabricated packing for waste packages for the Basalt Waste Isolation Project (BWIP). The principal functions of the packing are to minimize container corrosion during the 300 to 1000 years following repository closure and provide long-term control of the release of radionuclides from the waste package. The investigative work, discussed in this report, was specifically conceived to develop the design criteria for production of full-scale prototypical packing rings. The investigative work included the preparation of procedures, the preparation of fabrication materials, physical properties, and the determination of the engineering properties. The principal activities were the preparation of the materials and the determination of the physical properties. 21 refs., 20 figs., 14 tabs

Health physics, safety and medical services report for 1989

International Nuclear Information System (INIS)

Burt, A.K.; Bird, R.W.

1990-09-01

The Health Physics, Safety and Medical Services Report for Harwell Laboratory for 1989 includes data on the monitoring of the working environment, personnel monitoring, radiological incidents, disposal of radioactive waste and protection of the public. Work on emergency planning, non-radiological health and safety, occupational hygiene, operations support is also discussed. Finally the medical services available and the medical examinations performed are described. (UK)

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

International Nuclear Information System (INIS)

Thiesen, B.P.

1993-01-01

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

Waste Receiving and Processing (WRAP) Weight Scale Analysis Results

International Nuclear Information System (INIS)

JOHNSON, M.D.

2000-01-01

Fairbanks Weight Scales are used at the Waste Receiving and Processing (WRAP) facility to determine the weight of waste drums as they are received, processed, and shipped. Due to recent problems, discovered during calibration, the WRAP Engineering Department has completed this document which outlines both the investigation of the infeed conveyor scale failure in September of 1999 and recommendations for calibration procedure modifications designed to correct deficiencies in the current procedures

The public visits a nuclear waste site: Survey results from the West Valley Demonstration Project

International Nuclear Information System (INIS)

Hoffman, W.D.

1987-01-01

This paper discusses the results of the 1986 survey taken at the West Valley Demonstration Project Open House where a major nuclear waste cleanup is in progress. Over 1400 people were polled on what they think is most effective in educating the public on nuclear waste. A demographic analysis describes the population attending the event and their major interests in the project. Responses to attitudinal questions are examined to evaluate the importance of radioactive waste cleanup as an environmental issue and a fiscal responsibility. Additionally, nuclear power is evaluated on its public perception as an energy resource. The purpose of the study is to find out who visits a nuclear waste site and why, and to measure their attitudes on nuclear issues

Chemical compatibility screening results of plastic packaging to mixed waste simulants

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1995-01-01

We have developed a chemical compatibility program for evaluating transportation packaging components for transporting mixed waste forms. We have performed the first phase of this experimental program to determine the effects of simulant mixed wastes on packaging materials. This effort involved the screening of 10 plastic materials in four liquid mixed waste simulants. The testing protocol involved exposing the respective materials to ∼3 kGy of gamma radiation followed by 14 day exposures to the waste simulants of 60 C. The seal materials or rubbers were tested using VTR (vapor transport rate) measurements while the liner materials were tested using specific gravity as a metric. For these tests, a screening criteria of ∼1 g/m 2 /hr for VTR and a specific gravity change of 10% was used. It was concluded that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only VITON passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture simulant mixed waste, none of the seal materials met the screening criteria. It is anticipated that those materials with the lowest VTRs will be evaluated in the comprehensive phase of the program. For specific gravity testing of liner materials the data showed that while all materials with the exception of polypropylene passed the screening criteria, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals

The influence of slaughterhouse waste on fermentative H2 production from food waste: preliminary results.

Science.gov (United States)

Boni, Maria Rosaria; Sbaffoni, Silvia; Tuccinardi, Letizia

2013-06-01

The aim of this study was to evaluate the influence of slaughterhouse waste (SHW; essentially the skin, fats, and meat waste of pork, poultry, and beef) in a fermentative co-digestion process for H2 production from pre-selected organic waste taken from a refectory (food waste [FW]). Batch tests under mesophilic conditions were conducted in stirred reactors filled with different proportions of FW and SHW. The addition of 60% and 70% SHW to a mixture of SHW and FW improved H2 production compared to that in FW only, reaching H2-production yields of 145 and 109 ml g VS 0(-1), respectively, which are 1.5-2 times higher than that obtained with FW alone. Although the SHW ensured a more stable fermentative process due to its high buffering capacity, a depletion of H2 production occurred when SHW fraction was higher than 70%. Above this percentage, the formation of foam and aggregated material created non-homogenous conditions of digestion. Additionally, the increasing amount of SHW in the reactors may lead to an accumulation of long chain fatty acids (LCFAs), which are potentially toxic for anaerobic microorganisms and may inhibit the normal evolution of the fermentative process. Copyright © 2013 Elsevier Ltd. All rights reserved.

Techniques for improving shuffler assay results for 55-gallon waste drums

International Nuclear Information System (INIS)

Rinard, P.M.; Prettyman, T.H.; Stuenkel, D.

1994-01-01

Accurate assays of the fissile contents in waste drums are needed to ensure the most proper and economical handling and disposal of the waste. An improvement of accuracy will mean fewer drums disposed as transuranic waste when they really contain low-level waste, saving both money and burial sites. Shufflers are used for assaying waste drums and are very accurate with nonmoderating matrices (such as iron). In the active mode they count delayed neutrons released after fissions are induced by irradiation neutrons from a 252 Cf source. However, as the hydrogen density from matrices such as paper or gloves increases, the accuracy can suffer without proper attention. The neutron transport and fission probabilities change with the hydrogen density, causing the neutron count rate to vary with the position of the fissile material within the drum. The magnitude of this variation grows with the hydrogen density

Risk assessment on hazards for decommissioning safety of a nuclear facility

International Nuclear Information System (INIS)

Jeong, Kwan-Seong; Lee, Kune-Woo; Lim, Hyeon-Kyo

2010-01-01

A decommissioning plan should be followed by a qualitative and quantitative safety assessment of it. The safety assessment of a decommissioning plan is applied to identify the potential (radiological and non-radiological) hazards and risks. Radiological and non-radiological hazards arise during decommissioning activities. The non-radiological or industrial hazards to which workers are subjected during a decommissioning and dismantling process may be greater than those experienced during an operational lifetime of a facility. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities and as well as during accidents. The risk assessment method was developed by using risk matrix and fuzzy inference logic, on the basis of the radiological and non-radiological hazards for a decommissioning safety of a nuclear facility. Fuzzy inference of radiological and non-radiological hazards performs a mapping from radiological and non-radiological hazards to risk matrix. Defuzzification of radiological and non-radiological hazards is the conversion of risk matrix and priorities to the maximum criterion method and the mean criterion method. In the end, a composite risk assessment methodology, to rank the risk level on radiological and non-radiological hazards of the decommissioning tasks and to prioritize on the risk level of the decommissioning tasks, by simultaneously combining radiological and non-radiological hazards, was developed.

The influence of slaughterhouse waste on fermentative H{sub 2} production from food waste: Preliminary results

Energy Technology Data Exchange (ETDEWEB)

Boni, Maria Rosaria; Sbaffoni, Silvia; Tuccinardi, Letizia, E-mail: letizia.tuccinardi@uniroma1.it

2013-06-15

Highlights: • Co-digestion process finalized to bio-H{sub 2} production was tested in batch tests. • Slaughterhouse waste (SHW) and food waste (FW) were co-digested in different proportions. • The presence of SHW affected the H{sub 2} production from FW. • When SHW ranging between 50% and 70% the H{sub 2} production is improved. • SHW percentages above 70%, led to a depletion in H{sub 2} production. - Abstract: The aim of this study was to evaluate the influence of slaughterhouse waste (SHW; essentially the skin, fats, and meat waste of pork, poultry, and beef) in a fermentative co-digestion process for H{sub 2} production from pre-selected organic waste taken from a refectory (food waste [FW]). Batch tests under mesophilic conditions were conducted in stirred reactors filled with different proportions of FW and SHW. The addition of 60% and 70% SHW to a mixture of SHW and FW improved H{sub 2} production compared to that in FW only, reaching H{sub 2}-production yields of 145 and 109 ml gVS{sub 0}{sup -1}, respectively, which are 1.5–2 times higher than that obtained with FW alone. Although the SHW ensured a more stable fermentative process due to its high buffering capacity, a depletion of H{sub 2} production occurred when SHW fraction was higher than 70%. Above this percentage, the formation of foam and aggregated material created non-homogenous conditions of digestion. Additionally, the increasing amount of SHW in the reactors may lead to an accumulation of long chain fatty acids (LCFAs), which are potentially toxic for anaerobic microorganisms and may inhibit the normal evolution of the fermentative process.

Results from five years of treatability studies using hydraulic binders to stabilize low-level mixed waste at the INEL

International Nuclear Information System (INIS)

Gering, K.L.; Schwendiman, G.L.

1997-01-01

This paper summarizes work involving bench-scale solidification of nonincinerable, land disposal restricted low-level mixed waste. Waste forms included liquids, sludges, and solids; treatment techniques included hydraulic systems (Portland cement with and without additives), proprietary commercial formulations, and sulphur polymer cement. Solidification was performed to immobilize hazardous heavy metals (including mercury, lead, chromium, and cadmium), and volatile and semivolatile organic compounds. Pretreatment options for mixed wastes are discussed, using a decision tree based on the form of mixed waste and the type of hazardous constituents. Hundreds of small concrete monoliths were formed for a variety of waste types. The experimental parameters used for the hydraulic concrete systems include the ratio of waste to dry binder (Portland cement, proprietary materials, etc.), the total percentage of water in concrete, and the amount of concrete additives. The only parameter that was used for the sulfur polymer-based monoliths is ratio of waste to binder. Optimum concrete formulations or open-quotes recipesclose quotes for a given type of waste were derived through this study, as based on results from the Toxicity Characteristic Leaching Procedure analyses and a free liquids test. Overall results indicate that high waste loadings in the concrete can be achieved while the monolithic mass maintains excellent resistance to leaching of heavy metals. In our study the waste loadings in the concrete generally fell within the range of 0.5 to 2.0 kg mixed waste per kg dry binder. Likewise, the most favorable amount of water in concrete, which is highly dependent upon the concrete constituents, was determined to be generally within the range of 300 to 330 g/kg (30-33% by weight). The results of this bench-scale study will find applicability at facilities where mixed or hazardous waste solidification is a planned or ongoing activity. 19 refs., 1 fig., 5 tabs

Safety assessment of Novi Han radioactive waste repository - features, problems, results and perspectives

International Nuclear Information System (INIS)

Mateeva, M.

2000-01-01

This paper summarizes the work done and the achievements reached in the Novi Han radioactive waste repository safety assessment within the IAEA Model Project 'Increasing the safety of Novi Han radioactive waste repository BUL 4/005'. The overall safety assessment has a wide context, but the work reported here relates only to some details and results concerning the development and implementation of the appropriate methodology approach, model and computer code used for the calculations. Different steps and procedures are included for a better practical understanding of the obtained results during the safety assessment performance. The methodology approach is widely based on an international experience in safety analysis and implemented for evaluation computer code AMBER, which is one of the recommended from the safety assessments experts. (author)

The role and the results of the European Community's R and D work on radioactive waste management

International Nuclear Information System (INIS)

Orlowski, S.; Girardi, F.

1986-01-01

The role and results of the European Community's research and development (R and D) work on radioactive waste management are described. The R and D work includes: radioactive waste conditioning, characterization and storage, materials science studies for the storage, geological media confinement studies, and radionuclide migration investigations. Financial management and the long term, and the socio-political aspects of waste management, are also discussed. (U.K.)

Savannah River Plant low-level waste incinerator: Operational results and technical development

International Nuclear Information System (INIS)

Irujo, M.J.; Bucci, J.R.

1987-04-01

Volume reduction of solid and liquid low-level waste has been demonstrated at the Savannah River Plant (SRP) in the Waste Management Beta-Gamma Incinerator facility (BGI). The BGI uses a two-stage, controlled-air incinerator capable of processing 180 kg/hr (400 lbs/hr) of solid waste or 150 liters/hr (40 gal/hr) of liquid waste. These wastes are pyrolyzed in a substoichiometric air environment at 900 to 1100 degrees Celsius in the primary chamber. Products of partial combustion from the primary chamber are oxidized at 950 to 1150 degrees Celsius in the secondary chamber. A spray dryer, baghouse,and HEPA filter unit cool and filter the incinerator offgases. 2 refs., 9 tabs

TRU Waste Sampling Program: Volume I. Waste characterization

International Nuclear Information System (INIS)

Clements, T.L. Jr.; Kudera, D.E.

1985-09-01

Volume I of the TRU Waste Sampling Program report presents the waste characterization information obtained from sampling and characterizing various aged transuranic waste retrieved from storage at the Idaho National Engineering Laboratory and the Los Alamos National Laboratory. The data contained in this report include the results of gas sampling and gas generation, radiographic examinations, waste visual examination results, and waste compliance with the Waste Isolation Pilot Plant-Waste Acceptance Criteria (WIPP-WAC). A separate report, Volume II, contains data from the gas generation studies

Processing results of 1800 gallons of mercury and radioactively contaminated mixed waste rinse solution

International Nuclear Information System (INIS)

Thiesen, B.P.

1993-01-01

Mercury-contaminated rinse solution was successfully treated at the Idaho National Engineering Laboratory. This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 reactor shield tank. Approximately 6.8 m 3 (1,800 pi) of waste was generated and placed into 33 drums. Each drum contained precipitated sludge material ranging from 2--5 cm in depth, with the average depth of about 6 cm. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/mL while the average sludge contamination was about 13,800 pCi/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. The resulting solution after treatment had mercury levels at 0.0186 mg/l and radioactivity of 0.282 pCi/ml

60-Day waste compatibility safety issues and final results for AY-102 grab samples

Energy Technology Data Exchange (ETDEWEB)

Nuzum, J.L.

1997-01-31

Four grab samples (2AY-96-15, 2AY-96-16, 2AY-96-17, and 2AY-96-18) were taken from Riser 15D of Tank 241-AY-102 on October 8, 1996, and received by 222-S Laboratory on October 8, 1996. These samples were analyzed in accordance with Compatibility Grab Sampling and Analysis Plan (TSAP) and Data Quality Objectives for Tank Farms Waste Compatibility Program (DQO) in support of the Waste Compatibility Program. No notifications were required based on sample results.

Waste Preparation and Transport Chemistry: Results of the FY 2002 Studies

Energy Technology Data Exchange (ETDEWEB)

Hunt, R.D.

2003-07-10

The initial step in the remediation of nuclear waste stored at Hanford and the Savannah River Site (SRS) involves the retrieval and transfer of the waste to another tank or to a treatment facility. The retrieved waste can range from a filtered supernatant to a slurry. Nearly all of the recent solid formation problems encountered during waste transfers and subsequent treatment steps have involved decanted or filtered supernatants. Problems with slurry transfers have not yet surfaced, because tank farm operations at Hanford and the SRS have focused primarily on supernatant transfers and treatment. For example, the interim stabilization program at Hanford continues to reduce the level of supernatants and interstitial liquids in its single-shell tanks through saltwell pumping of filtered liquid. In addition, at present, the cross-site transfer lines at Hanford can be used only to transfer liquids. Another reason for fewer problems with slurry transfers involves the additions of large quantities of dilution water prior to the transfer. When the waste is transferred, a drop in temperature is expected because most transfer lines are not heated. However, the dilution water reduces or eliminates solid formation caused by this temperature drop. In sharp contrast, decanted or filtered supernatants are near or at saturation for certain compounds. In such cases, tank farm operators must continue to evaporate their liquid waste since available tank space is quite limited. Solid formation can occur when the temperature of saturated solutions drops even slightly. The evaporation step can also lead to the formation of problematic solids. At the SRS, the evaporation of a relatively dilute waste stream was suspended due to the formation of deposits in the evaporator system. Therefore, small drops in temperature or evaporation can lead to problematic solid formations.

Waste Preparation and Transport Chemistry: Results of the FY 2002 Studies

International Nuclear Information System (INIS)

Hunt, R.D.

2003-01-01

The initial step in the remediation of nuclear waste stored at Hanford and the Savannah River Site (SRS) involves the retrieval and transfer of the waste to another tank or to a treatment facility. The retrieved waste can range from a filtered supernatant to a slurry. Nearly all of the recent solid formation problems encountered during waste transfers and subsequent treatment steps have involved decanted or filtered supernatants. Problems with slurry transfers have not yet surfaced, because tank farm operations at Hanford and the SRS have focused primarily on supernatant transfers and treatment. For example, the interim stabilization program at Hanford continues to reduce the level of supernatants and interstitial liquids in its single-shell tanks through saltwell pumping of filtered liquid. In addition, at present, the cross-site transfer lines at Hanford can be used only to transfer liquids. Another reason for fewer problems with slurry transfers involves the additions of large quantities of dilution water prior to the transfer. When the waste is transferred, a drop in temperature is expected because most transfer lines are not heated. However, the dilution water reduces or eliminates solid formation caused by this temperature drop. In sharp contrast, decanted or filtered supernatants are near or at saturation for certain compounds. In such cases, tank farm operators must continue to evaporate their liquid waste since available tank space is quite limited. Solid formation can occur when the temperature of saturated solutions drops even slightly. The evaporation step can also lead to the formation of problematic solids. At the SRS, the evaporation of a relatively dilute waste stream was suspended due to the formation of deposits in the evaporator system. Therefore, small drops in temperature or evaporation can lead to problematic solid formations

Mechanisms of gas retention and release: Experimental results for Hanford waste tanks 241-AW-101 and 241-AN-103

Energy Technology Data Exchange (ETDEWEB)

Rassat, S.D.; Gauglitz, P.A.; Bredt, P.R.; Mahoney, L.A.; Forbes, S.V.; Tingey, S.M.

1997-09-01

The 177 storage tanks at Hanford contain a vast array of radioactive waste forms resulting, primarily, from nuclear materials processing. Through radiolytic, thermal, and other decomposition reactions of waste components, gaseous species including hydrogen, ammonia, and the oxidizer nitrous oxide are generated within the waste tanks. Many of these tanks are known to retain and periodically release quantities of these flammable gas mixtures. The primary focus of the Flammable Gas Project is the safe storage of Hanford tank wastes. To this end, we strive to develop an understanding of the mechanisms of flammable gas retention and release in Hanford tanks through laboratory investigations on actual tank wastes. These results support the closure of the Flammable Gas Unreviewed Safety Question (USQ) on the safe storage of waste tanks known to retain flammable gases and support resolution of the broader Flammable Gas Safety Issue. The overall purpose of this ongoing study is to develop a comprehensive and thorough understanding of the mechanisms of flammable gas retention and release. The first objective of the current study was to classify bubble retention and release mechanisms in two previously untested waste materials from Tanks 241-AN-103 (AN-103) and 241-AW-101 (AW-101). Results were obtained for retention mechanisms, release characteristics, and the maximum gas retention. In addition, unique behavior was also documented and compared with previously studied waste samples. The second objective was to lengthen the duration of the experiments to evaluate the role of slowing bubble growth on the retention and release behavior. Results were obtained for experiments lasting from a few hours to a few days.

Mechanisms of gas retention and release: Experimental results for Hanford waste tanks 241-AW-101 and 241-AN-103

International Nuclear Information System (INIS)

Rassat, S.D.; Gauglitz, P.A.; Bredt, P.R.; Mahoney, L.A.; Forbes, S.V.; Tingey, S.M.

1997-09-01

The 177 storage tanks at Hanford contain a vast array of radioactive waste forms resulting, primarily, from nuclear materials processing. Through radiolytic, thermal, and other decomposition reactions of waste components, gaseous species including hydrogen, ammonia, and the oxidizer nitrous oxide are generated within the waste tanks. Many of these tanks are known to retain and periodically release quantities of these flammable gas mixtures. The primary focus of the Flammable Gas Project is the safe storage of Hanford tank wastes. To this end, we strive to develop an understanding of the mechanisms of flammable gas retention and release in Hanford tanks through laboratory investigations on actual tank wastes. These results support the closure of the Flammable Gas Unreviewed Safety Question (USQ) on the safe storage of waste tanks known to retain flammable gases and support resolution of the broader Flammable Gas Safety Issue. The overall purpose of this ongoing study is to develop a comprehensive and thorough understanding of the mechanisms of flammable gas retention and release. The first objective of the current study was to classify bubble retention and release mechanisms in two previously untested waste materials from Tanks 241-AN-103 (AN-103) and 241-AW-101 (AW-101). Results were obtained for retention mechanisms, release characteristics, and the maximum gas retention. In addition, unique behavior was also documented and compared with previously studied waste samples. The second objective was to lengthen the duration of the experiments to evaluate the role of slowing bubble growth on the retention and release behavior. Results were obtained for experiments lasting from a few hours to a few days

Environmental monitoring for EG and G Idaho facilities at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Tkachyk, J.W.; Wright, K.C.; Wilhelmsen, R.N.

1990-08-01

This report describes the 1989 environmental-monitoring activities of the Environmental Monitoring Unit of EG ampersand G Idaho, Inc., at EG ampersand G-operated facilities at the Idaho National Engineering Laboratory (INEL). The major facilities monitored include the Radioactive Waste Management Complex, the Waste Experimental Reduction Facility, the Mixed Waste Storage Facility, and two surplus facilities. Additional monitoring activities performed by Environmental Monitoring are also discussed, including drinking-water monitoring and nonradiological liquid-effluent monitoring, as well as data management. The primary purposes of monitoring are to evaluate environmental conditions and to provide and interpret data, in compliance with applicable regulations, to ensure protection of human health and the environment. This report compares 1989 environmental-monitoring data with derived concentration guides and with data from previous years. This report also presents results of sampling performed by the Radiological and Environmental Sciences Laboratory and by the United States Geological Survey. 17 refs., 49 figs., 11 tabs

The extent of food waste generation across EU-27: different calculation methods and the reliability of their results.

Science.gov (United States)

Bräutigam, Klaus-Rainer; Jörissen, Juliane; Priefer, Carmen

2014-08-01

The reduction of food waste is seen as an important societal issue with considerable ethical, ecological and economic implications. The European Commission aims at cutting down food waste to one-half by 2020. However, implementing effective prevention measures requires knowledge of the reasons and the scale of food waste generation along the food supply chain. The available data basis for Europe is very heterogeneous and doubts about its reliability are legitimate. This mini-review gives an overview of available data on food waste generation in EU-27 and discusses their reliability against the results of own model calculations. These calculations are based on a methodology developed on behalf of the Food and Agriculture Organization of the United Nations and provide data on food waste generation for each of the EU-27 member states, broken down to the individual stages of the food chain and differentiated by product groups. The analysis shows that the results differ significantly, depending on the data sources chosen and the assumptions made. Further research is much needed in order to improve the data stock, which builds the basis for the monitoring and management of food waste. © The Author(s) 2014.

Performance test results of noninvasive characterization of RCRA surrogate waste by prompt gamma neutron activation analysis

International Nuclear Information System (INIS)

Gehrke, R.J.; Propp, W.A.

1997-11-01

A performance evaluation to determine the feasibility of using prompt gamma neutron activation analysis (PGNAA) for noninvasive, quantitative assay of mixed waste containers was sponsored by DOE's Office of Technology Development (OTD), the Mixed Waste Focus Area (MWFA), and the Idaho National Engineering and Environmental Laboratory (INEEL). The evaluation was conducted using a surrogate waste, based on Portland cement, that was spiked with three RCRA metals, mercury, cadmium, and lead. The results indicate that PGNAA has potential as a process monitor. However, further development is required to improve its sensitivity to meet regulatory requirements for determination of these RCRA metals

Solid Waste Management Requirements Definition for Advanced Life Support Missions: Results

Science.gov (United States)

Alazraki, Michael P.; Hogan, John; Levri, Julie; Fisher, John; Drysdale, Alan

2002-01-01

Prior to determining what Solid Waste Management (SWM) technologies should be researched and developed by the Advanced Life Support (ALS) Project for future missions, there is a need to define SWM requirements. Because future waste streams will be highly mission-dependent, missions need to be defined prior to developing SWM requirements. The SWM Working Group has used the mission architecture outlined in the System Integration, Modeling and Analysis (SIMA) Element Reference Missions Document (RMD) as a starting point in the requirement development process. The missions examined include the International Space Station (ISS), a Mars Dual Lander mission, and a Mars Base. The SWM Element has also identified common SWM functionalities needed for future missions. These functionalities include: acceptance, transport, processing, storage, monitoring and control, and disposal. Requirements in each of these six areas are currently being developed for the selected missions. This paper reviews the results of this ongoing effort and identifies mission-dependent resource recovery requirements.

TSA waste stream and final waste form composition

International Nuclear Information System (INIS)

Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

1993-01-01

A final vitrified waste form composition, based upon the chemical compositions of the input waste streams, is recommended for the transuranic-contaminated waste stored at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The quantities of waste are large with a considerable uncertainty in the distribution of various waste materials. It is therefore impractical to mix the input waste streams into an ''average'' transuranic-contaminated waste. As a result, waste stream input to a melter could vary widely in composition, with the potential of affecting the composition and properties of the final waste form. This work examines the extent of the variation in the input waste streams, as well as the final waste form under conditions of adding different amounts of soil. Five prominent Rocky Flats Plant 740 waste streams are considered, as well as nonspecial metals and the ''average'' transuranic-contaminated waste streams. The metals waste stream is the most extreme variation and results indicate that if an average of approximately 60 wt% of the mixture is soil, the final waste form will be predominantly silica, alumina, alkaline earth oxides, and iron oxide. This composition will have consistent properties in the final waste form, including high leach resistance, irrespective of the variation in waste stream. For other waste streams, much less or no soil could be required to yield a leach resistant waste form but with varying properties

Mineralogical conversion of asbestos containing materials

International Nuclear Information System (INIS)

Pulsford, S.K.; Foltz, A.D.; Ek, R.B.

1996-01-01

The principal objective of the Technical Task Plan (TTP) is to demonstrate a thermal-chemical mineralogical asbestos conversion unit at the Hanford Site, which converts non-radiological asbestos containing materials (ACMs) into an asbestos-free material. The permanent thermal-chemical mineralogical conversion of ACMs to a non-toxic, non-hazardous, potentially marketable end product should not only significantly reduce the waste stream volumes but terminate the open-quotes cradle to graveclose quotes ownership liabilities

Status of test results of electrochemical organic oxidation of a tank 241-SY-101 simulated waste

International Nuclear Information System (INIS)

Colby, S.A.

1994-06-01

This report presents scoping test results of an electrochemical waste pretreatment process to oxidize organic compounds contained in the Hanford Site's radioactive waste storage tanks. Electrochemical oxidation was tested on laboratory scale to destroy organics that are thought to pose safety concerns, using a nonradioactive, simulated tank waste. Minimal development work has been applied to alkaline electrochemical organic destruction. Most electrochemical work has been directed towards acidic electrolysis, as in the metal purification industry, and silver catalyzed oxidation. Alkaline electrochemistry has traditionally been associated with the following: (1) inefficient power use, (2) electrode fouling, and (3) solids handling problems. Tests using a laboratory scale electrochemical cell oxidized surrogate organics by applying a DC electrical current to the simulated tank waste via anode and cathode electrodes. The analytical data suggest that alkaline electrolysis oxidizes the organics into inorganic carbonate and smaller carbon chain refractory organics. Electrolysis treats the waste without adding chemical reagents and at ambient conditions of temperature and pressure. Cell performance was not affected by varying operating conditions and supplemental electrolyte additions

The Behaviours of Cementitious Materials in Long Term Storage and Disposal of Radioactive Waste. Results of a Coordinated Research Project

International Nuclear Information System (INIS)

2013-09-01

Radioactive waste with widely varying characteristics is generated from the operation and maintenance of nuclear power plants, nuclear fuel cycle facilities, research laboratories and medical facilities. This waste must be treated and conditioned, as necessary, to provide waste forms acceptable for safe storage and disposal. Many countries use cementitious materials (concrete, mortar, etc.) as a containment matrix for immobilization, as well as for engineered structures of disposal facilities. Radionuclide release is dependent on the physicochemical properties of the waste forms and packages, and on environmental conditions. In the use of cement, the diffusion process and metallic corrosion can induce radionuclide release. The advantage of cementitious materials is the added stability and mechanical support during storage and disposal of waste. Long interim storage is becoming an important issue in countries where it is difficult to implement low level waste and intermediate level waste disposal facilities, and in countries where cement is used in the packaging of waste that is not suitable for shallow land disposal. This coordinated research project (CRP), involving 24 research organizations from 21 Member States, investigated the behaviour and performance of cementitious materials used in an overall waste conditioning system based on the use of cement - including waste packaging (containers), waste immobilization (waste form) and waste backfilling - during long term storage and disposal. It also considered the interactions and interdependencies of these individual elements (containers, waste, form, backfill) to understand the processes that may result in degradation of their physical and chemical properties. The main research outcomes of the CRP are summarized in this report under four topical sections: (i) conventional cementitious systems; (ii) novel cementitious materials and technologies; (iii) testing and waste acceptance criteria; and (iv) modelling long

The Behaviours of Cementitious Materials in Long Term Storage and Disposal of Radioactive Waste. Results of a Coordinated Research Project

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-09-15

Radioactive waste with widely varying characteristics is generated from the operation and maintenance of nuclear power plants, nuclear fuel cycle facilities, research laboratories and medical facilities. This waste must be treated and conditioned, as necessary, to provide waste forms acceptable for safe storage and disposal. Many countries use cementitious materials (concrete, mortar, etc.) as a containment matrix for immobilization, as well as for engineered structures of disposal facilities. Radionuclide release is dependent on the physicochemical properties of the waste forms and packages, and on environmental conditions. In the use of cement, the diffusion process and metallic corrosion can induce radionuclide release. The advantage of cementitious materials is the added stability and mechanical support during storage and disposal of waste. Long interim storage is becoming an important issue in countries where it is difficult to implement low level waste and intermediate level waste disposal facilities, and in countries where cement is used in the packaging of waste that is not suitable for shallow land disposal. This coordinated research project (CRP), involving 24 research organizations from 21 Member States, investigated the behaviour and performance of cementitious materials used in an overall waste conditioning system based on the use of cement - including waste packaging (containers), waste immobilization (waste form) and waste backfilling - during long term storage and disposal. It also considered the interactions and interdependencies of these individual elements (containers, waste, form, backfill) to understand the processes that may result in degradation of their physical and chemical properties. The main research outcomes of the CRP are summarized in this report under four topical sections: (i) conventional cementitious systems; (ii) novel cementitious materials and technologies; (iii) testing and waste acceptance criteria; and (iv) modelling long

Waste Tank Vapor Program: Vapor space characterization of waste tank 241-T-111. Results from samples collected on January 20, 1995

International Nuclear Information System (INIS)

Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.; Pool, K.H.; McVeety, B.D.; Olsen, K.B.; Bredt, O.P.; Fruchter, J.S.; Goheen, S.C.

1995-10-01

This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-T-111. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

Waste Tank Vapor Program: Vapor space characterization of waste tank 241; C-102: Results from samples collected on August 23, 1994

International Nuclear Information System (INIS)

Klinger, G.S.; Clauss, T.W.; Ligotke, M.W.

1995-10-01

This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-C-102. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

Vapor space characterization of waste tank 241-U-111: Results from samples collected on February 28, 1995. Waste Tank Vapor Program

International Nuclear Information System (INIS)

Clauss, T.W.; Pool, K.H.; McVeety, B.D.; Bredt, O.P.; Goheen, S.C.; Ligotke, M.W.; Lucke, R.B.; Klinger, G.S.; Fruchter, J.S.

1995-07-01

This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-U-111. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

Treatment and conditioning of low-level radioactive waste in Belgium: initial operating results of the Cilva facility

International Nuclear Information System (INIS)

Monsch, O.; Renard, C.; Deckers, J.; Luycx, P.

1995-01-01

The Belgian National Radioactive Waste and Enriched Fissile Material Agency (ONDRAF), which is responsible for the management of all radioactive waste in Belgium, recently decided to commission the CILVA facility. Operation of this facility, which comprises a number of units for the treatment of low-level radwaste, has been contracted to ONDRAF's Belgoprocess subsidiary based at the Dessel site. A consortium comprising SGN and Fabricom was in charge of building the CILVA facility's waste preparation and conditioning (concrete solidification) units. The concrete solidification processes, which were devised and developed by SGN, have been qualified to secure ONDRAF certification of the process and the facility. This enabled active commissioning of the waste conditioning unit in mid-August 1994. Active commissioning of the waste preparation unit was carried out in several stages up to the beginning of 1995 in accordance with operating requirements. Initial operating results of the two units are presented. (author)

Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

International Nuclear Information System (INIS)

Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

2011-01-01

This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

2011-12-01

This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

Viability Assessment of a Repository at Yucca Mountain. Volume 2: Preliminary Design Concept for the Repository and Waste Package

Energy Technology Data Exchange (ETDEWEB)

None

1998-12-01

This volume describes the major design features of the Monitored Geologic Repository. This document is not intended to provide an exhaustive, detailed description of the repository design. Rather, this document summarizes the major systems and primary elements of the design that are radiologically significant, and references the specific technical documents and design analyses wherein the details can be found. Not all portions of the design are at the same level of completeness. Highest priority has been given to assigning resources to advance the design of the Monitored Geologic Repository features that are important to radiological safety and/or waste isolation and for which there is no NRC licensing precedent. Those features that are important to radiological safety and/or waste isolation, but for which there is an NRC precedent, receive second priority. Systems and features that have no impact on radiological safety or waste isolation receive the lowest priority. This prioritization process, referred to as binning, is discussed in more detail in Section 2.3. Not every subject discussed in this volume is given equal treatment with regard to the level of detail provided. For example, less detail is provided for the surface facility design than for the subsurface and waste package designs. This different level of detail is intentional. Greater detail is provided for those functions, structures, systems, and components that play key roles with regard to protecting radiological health and safety and that are not common to existing nuclear facilities already licensed by NRC. A number of radiological subjects are not addressed in the VA, (e.g., environmental qualification of equipment). Environmental qualification of equipment and other radiological safety considerations will be addressed in the LA. Non-radiological safety considerations such as silica dust control and other occupational safety considerations are considered equally important but are not addressed in

Quarry waste management and recovery: first results connected to Carrara marble ravaneti (Italy)

Science.gov (United States)

Antonella Dino, Giovanna; Chiappino, Claudia; Rossetti, Piergiorgio

2017-04-01

Quarry waste (QW) represents a huge economic and environmental issue, due to loss of resources and to economic and environmental costs connected to waste management and landfilling activities. In many cases, valuable Raw Materials (RM) and Secondary Raw Materials (SRM) can be supplied by enhancing the QW recovery. In Italy large amounts of QW have been and still are dumped: such materials, if their quality (chemical, mineralogical, physical characteristics) and quantity are adequate, and if the impacts connected to their management are positive, can represent a valuable resource for SRM exploitation. Several dimension stone quarries have been and are interested by researches as for QW exploitation. Some researches show positive results, which are the basis for QW recovery (both from waste streams and from quarry dumps exploitation): a noticeable example is represented by Carrara marble waste. The Carrara quarry basin is characterized by ca. one hundred quarries for colored and white marble exploitation. The waste production can be summarized in: 80 Mm3 waste present in old quarry dumps (Ravaneti) and 3 Mm3/y of waste stream from quarrying activities. At present only 0.5 Mm3/y of QW is exploited for SRM production, causing a huge loss of resource. This has been the background for a preliminary research, on Carrara marble Ravaneti characterization, which was carried out thanks to the close cooperation between University of Torino, Società Apuana Marmi srl, and SET srl. In 2015, two QW dumping areas, Calocara and Lorano, were selected as representative for sampling activities. Three main sample categories were individuated based on granulometry (0.5-4 mm, 0-25 mm, 0-150 mm) to be characterized (size distribution, density, Atterberg limits, Los Angeles test, freezing and heat tests, flat and shape indexes, geochemistry, mineralogy). The results obtained are promising: the physical characterization shows an attitude for Carrara QW to be recovered as crushed materials

Removal of strontium and transuranics from Hanford waste via hydrothermal processing -- FY 1994/95 test results

International Nuclear Information System (INIS)

Orth, R.J.; Schmidt, A.J.; Elmore, M.R.; Hart, T.R.; Neuenschwander, G.G.; Gano, S.R.; Lehmann, R.W.; Momont, J.A.

1995-09-01

Under the Tank Waste Remediation System (TWRS) Pretreatment Technology Development Project, Pacific Northwest Laboratory (PNL) is evaluating and developing organic destruction technologies that may be incorporated into the Initial Pretreatment Module (IPM) to treat Hanford tank waste. Organic (and ferrocyanide) destruction removes the compounds responsible for waste safety issues, and conditions the supernatant for low-level waste disposal by removing compounds that may be responsible for promoting strontium and transuranic (TRU) components solubility. Destruction or defunctionalization of complexing organics in tank wastes eliminates organic species that can reduce the efficiency of radionuclide (E.g., 90 Sr) separation processes, such as ion exchange, solvent extraction, and precipitation. The technologies being evaluated and tested for organic destruction are low-temperature hydrothermal processing (HTP) and wet air oxidation (WAO). Four activities are described: Batch HTP/WAO testing with Actual Tank Waste (Section 3.0), Batch HTP Testing with Simulant (Section 4.0), Batch WAO testing with Simulant (Section 5.0), and Continuous Bench-scale WAO Testing with Simulant (Section 6.0). For each of these activities, the objectives, test approach, results, status, and direction of future investigations are discussed. The background and history of the HTP/WAO technology is summarized below. Conclusions and Recommendations are provided in Section 2.0. A continuous HTP off-gas safety evaluation conducted in FY 1994 is included as Appendix A

Effects of container material on PCT leach test results for high-level nuclear waste glasses

International Nuclear Information System (INIS)

Xing, S.B.; Pegg, I.L.

1994-01-01

A glass-based waste form used for the immobilization of high-level nuclear wastes should exhibit good resistance to aqueous corrosion since typically this is the primary process by which radionucleides could be released into the environment upon failure of other barriers. In the USA, the Waste Acceptance Product Specifications (WAPS) provides a set of requirements to ensure the consistency of the waste forms produced and specifies the Product Consistency Test (PCT) as a measure of relative chemical durability. While the PCT procedure permits usage of both Teflon and stainless steel vessels for testing of simulated development glasses, Teflon is not permitted for testing of production glasses due to radiative degradation. The results presented in this paper indicate that there are very significant differences between tests conducted in the two types of vessels due to the well-known permeability of Teflon to atmospheric carbon dioxide which results in lowering of the solution pH and a consequent reduction in the leach rate of silicate glasses. A wide range of nuclear waste glass compositions was subjected to the PCT procedure using both Teflon and stainless steel vessels. The magnitude of the effect (up to a factor of four for B, Na, Li concentrations) depends strongly on glass composition, therefore the isolated checks performed previously were inconclusive. The permeability to CO, of two types of Teflon vessels specified in the PCT procedure was directly measured using buffer solutions: ingress of CO, is linear in time, strongly pH-dependent, and was as high as 100 ppm after 7 days. In actual PCT tests in Teflon vessels, the total CO, content was 560 ppm after 87 days and 1930 ppm after one year

Factor analysis on hazards for safety assessment in decommissioning workplace of nuclear facilities using a semantic differential method

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwan-Seong [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)], E-mail: ksjeongl@kaeri.re.kr; Lim, Hyeon-Kyo [Chungbuk National University, 410 Sungbong-ro, Heungduk-gu, Cheongju, Chungbuk 361-763 (Korea, Republic of)

2009-10-15

The decommissioning of nuclear facilities must be accomplished according to its structural conditions and radiological characteristics. An effective risk analysis requires basic knowledge about possible risks, characteristics of potential hazards, and comprehensive understanding of the associated cause-effect relationships within a decommissioning for nuclear facilities. The hazards associated with a decommissioning plan are important not only because they may be a direct cause of harm to workers but also because their occurrence may, indirectly, result in increased radiological and non-radiological hazards. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities as well as during accidents. Therefore, to prepare the safety assessment for decommissioning of nuclear facilities, the radiological and non-radiological hazards should be systematically identified and classified. With a semantic differential method of screening factor and risk perception factor, the radiological and non-radiological hazards are screened and identified.

ICPP [Idaho Chemical Processing Plant] environmental monitoring report, CY-1988

International Nuclear Information System (INIS)

Krivanek, K.R.

1989-08-01

Summarized in this report are the data collected through Environmental Monitoring programs conducted at the Idaho Chemical Processing Plant (ICPP) by the Environmental Engineering (EE) Section of the Nuclear and Industrial Safety (N and IS) Department. The ICPP is responsible for complying with all applicable Federal, State, Local and DOE Rules, Regulations and Orders. Radiological effluent and emissions are regulated by the DOE. The Environmental Protection Agency (EPA) regulates all nonradiological waste resulting from the ICPP operations including all airborne, liquid, and solid waste. The EE subsection completed a Quality Assurance (QA) Plan for Environmental Monitoring activities during the third quarter of 1986. QA activities have resulted in the ICPP's implementation of the Environmental Protection Agency rules and guidelines pertaining to the collection, analyses, and reporting of environmentally related samples. Where no approved methods for analyses existed for radionuclides, currently used methods were submitted for the EPA approval. 33 figs., 14 tabs

Radioactive waste discharges from UKAEA establishments during 1996 and associated monitoring results

International Nuclear Information System (INIS)

Morton, A.K.M.; Forbes, S.A.; Hughes, B.; Richardson, E.

1997-08-01

This annual report is published by the Safety Directorate of the United Kingdom Atomic Energy Authority (UKAEA) and provides information on radioactive discharges from its sites. The Culcheth site was closed and then redeveloped during the end of 1993 and the Springfields site became part of BNFL in October 1994. No operations involving the need to discharge radioactivity are undertaken at the Risley site. After discussions with the Authorising Departments at that time, the discharge authorisations were revoked on 1 July 1994. These sites are therefore no longer included in this report. UKAEA has published annual radioactive waste discharges and associated monitoring results since 1963. This report is intended to give a relatively short factual overview of UKAEA waste discharge and disposal, and its impact on the environment. Additional information may be found in annual discharge reports published by the individual UKAEA establishments and the UKAEA Report on Safety and the Environment 1996-97 due to be issued at the end of September 1997. (UK)

Long-term leaching behavior of simulated Savannah River Plant waste glass: Part 1, MCC-1 leachability results, four-year leaching data

International Nuclear Information System (INIS)

Wicks, G.G.; Stone, J.A.; Chandler, G.T.; Williams, S.

1986-08-01

Long-term leaching data were obtained on SRP 131/TDS waste glass using MCC-1 or slightly modified MCC-1 standard leaching tests. Experiments were conducted out to four years at 40 0 C and 3-1/2 years at 90 0 C. These experiments have produced the longest standardized leaching data currently available in the waste management community. Long-term leaching data provide important input to modeling of waste glass behavior and ultimate prediction of waste glass performance. In this study, the leaching behavior of SRP waste glass was found to be excellent; leachates based on a variety of elements were not only very low, but also improved with increasing time. In addition to these data, results are also reported from another independent Savannah River study. Leaching behavior at 40 0 C and 90 0 C was assessed not only for a similar SRP 131 waste glass composition, but also for extreme waste glass compositions involving high-iron and high-aluminum waste. In addition, these experiments were performed using not only a standard deionized water leachant, but also simplified brine and silicate groundwater simulations. These two large data bases will be summarized and correlated along with some of the more interesting results recently reported in another study, a two-year leaching program performed on a similar SRP waste glass composition at Battelle Pacific Northwest Laboratories

Response of soil microorganisms to radioactive oil waste: results from a leaching experiment

Science.gov (United States)

Galitskaya, P.; Biktasheva, L.; Saveliev, A.; Ratering, S.; Schnell, S.; Selivanovskaya, S.

2015-06-01

Oil wastes produced in large amounts in the processes of oil extraction, refining, and transportation are of great environmental concern because of their mutagenicity, toxicity, high fire hazardousness, and hydrophobicity. About 40% of these wastes contain radionuclides; however, the effects of oil products and radionuclides on soil microorganisms are frequently studied separately. The effects on various microbial parameters of raw waste containing 575 g of total petroleum hydrocarbons (TPH) kg-1 waste, 4.4 of 226Ra, 2.8 of 232Th, and 1.3 kBq kg-1 of 40K and its treated variant (1.6 g kg-1 of TPH, 7.9 of 226Ra, 3.9 of 232Th, and 183 kBq kg-1 of 40K) were examined in a leaching column experiment to separate the effects of hydrocarbons from those of radioactive elements. The raw waste sample (H) was collected from tanks during cleaning and maintenance, and a treated waste sample (R) was obtained from equipment for oil waste treatment. Thermal steam treatment is used in the production yard to reduce the oil content. The disposal of H waste samples on the soil surface led to an increase in the TPH content in soil: it became 3.5, 2.8, and 2.2 times higher in the upper (0-20 cm), middle (20-40 cm), and lower (40-60cm) layers, respectively. Activity concentrations of 226Ra and 232Th increased in soil sampled from both H- and R- columns in comparison to their concentrations in control soil. The activity concentrations of these two elements in samples taken from the upper and middle layers were much higher for the R-column compared to the H-column, despite the fact that the amount of waste added to the columns was equalized with respect to the activity concentrations of radionuclides. The H waste containing both TPH and radionuclides affected the functioning of the soil microbial community, and the effect was more pronounced in the upper layer of the column. Metabolic quotient and cellulase activity were the most sensitive microbial parameters as their levels were changed 5

Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004

International Nuclear Information System (INIS)

2007-10-01

This publication contains the results of an IAEA Coordinated Research Project (CRP). It provides a basis for understanding the potential interactions of waste form and repository environment, which is necessary for the development of the design and safety case for deep disposal. Types of high level waste matrices investigated include spent fuel, glasses and ceramics. Of particular interest are the experimental results pertaining to ceramic forms such as SYNROC. This publication also outlines important areas for future work, namely, standardized, collaborative experimental protocols for package-release studies, structured development and calibration of predictive models linking the performance of packaged waste and the repository environment, and studies of the long term behaviour of the wastes, including active waste samples

Mechanisms of gas retention and release: Experimental results for Hanford single-shell waste tanks 241-A-101, 241-S-106, and 241-U-103

International Nuclear Information System (INIS)

Rassat, S.D.; Caley, S.M.; Bredt, P.R.; Gauglitz, P.A.; Rinehart, D.E.; Forbes, S.V.

1998-09-01

The 177 underground waste storage tanks at the Hanford Site contain millions of gallons of radioactive waste resulting from the purification of nuclear materials and related processes. Through various mechanisms, flammable gas mixtures of hydrogen, ammonia, methane, and nitrous oxide are generated and retained in significant quantities within the waste in many (∼25) of these tanks. The potential for large releases of retained gas from these wastes creates a flammability hazard. It is a critical component of the effort to understand the flammability hazard and a primary goal of this laboratory investigation to establish an understanding of the mechanisms of gas retention and release in these wastes. The results of bubble retention experimental studies using waste samples from several waste tanks and a variety of waste types support resolution of the Flammable Gas Safety Issue. Gas bubble retention information gained in the pursuit of safe storage will, in turn, benefit future waste operations including salt-well pumping, waste transfers, and sluicing/retrieval

Actinide partitioning-transmutation program final report. VI. Short-term risk analysis of reprocessing, refabrication, and transportation: summary

International Nuclear Information System (INIS)

Fullwood, R.; Jackson, R.

1980-03-01

A Partitioning-Transmutation (PT) fuel cycle is being compared to a Reference cycle employing conventional fuel-material recovery methods. The PT cycle uses enhanced recovery methods so that most of the long-lived actinides are recycled to nuclear power plants and transmuted thereby reducing the waste toxicity. This report compares the two fuel cycles on the basis of the short-term radiological and nonradiological risks. The accidental radiological risk to the public is analyzed by estimating the probabilities of sets of accidents; the consequences are calculated using the risk, which is RAC code. Routine radiological risks to the public are estimated from the calculated release amounts, also using the CRAC code. Radiological occupational risks are determined from prior experience, projected standards, and estimates of accident risk. Nonradiological risks are calculated from the number of personnel involved, historical experience, and epidemiological studies. Result of this analysis is that the short-term risk of PT is 2.9 times greater than that of the Reference cycle, primarily due to the larger amount of industry. The nonradiological risk which is about 150 times greater than the radiological risk. If the radiological risk is consdered alone, the ratio of PT to Reference risk is 3, composed as follows: radiological operations affecting the public 5, radiological operations affecting the workers 1.7, and radiological accidents affecting the public 1.4, all in the order of decreasing risk. The absolute risk as estimated for the fuel cycle portions considered in this report is 0.91 fatality/GWe-year for the PT cycle and 0.34 fatality/GWe-year for the reference cycle; this compares with 1.5 for nuclear and 150 for coal. All of the risks assumed here are associated with the production of one billion watts of electricity (GWe) per year

Actinide partitioning-transmutation program final report. VI. Short-term risk analysis of reprocessing, refabrication, and transportation: appendix

International Nuclear Information System (INIS)

Fullwood, R.R.; Jackson, R.

1980-01-01

The Chemical Technology Division of the Oak Ridge National Laboratory has prepared a set of documents that evaluate a Partitioning-Transmutation (PT) fuel cycle relative to a Reference cycle employing conventional fuel-material recovery methods. The PT cycle uses enhanced recovery methods so that most of the long-lived actinides are recycled to nuclear power plants and transmuted to shorter-lived materials, thereby reducing the waste toxicity. This report compares the two fuel cycles on the basis of the short-term radiological and nonradiological risks they present to the public and to workers. The accidental radiological risk to the public is analyzed by estimating the probabilities of sets of accidents; the consequences are calculated using the CRAC code appropriately modified for the material composition. Routine radiological risks to the public are estimated from the calculated release amounts; the effects are calculated using the CRAC code. Radiological occupational risks are determined from prior experience, projected standards, and estimates of accident risk. Nonradiological risks are calculated from the number of personnel involved, historical experience, and epidemiological studies. The result of this analysis is that the short-term risk of PT is 2.9 times greater than that of the Reference cycle, primarily due to the larger amount of industry. This conclusion is strongly dominated by the nonradiological risk, which is about 150 times greater than the radiological risk. The absolute risk as estimated for the fuel cycle portions considered in this report is 0.91 fatalities/GWe-year for the PT cycle and 0.34 fatalities/GWe-year for the Reference cycle. This should be compared with Inhaber's estimate of 1.5 for nuclear and 150 for coal. All of the risks assumed here are associated with the production of one billion watts of electricity (GWe) per year

Actinide partitioning-transmutation program final report. VI. Short-term risk analysis of reprocessing, refabrication, and transportation: appendix

Energy Technology Data Exchange (ETDEWEB)

Fullwood, R.R.; Jackson, R.

1980-01-01

The Chemical Technology Division of the Oak Ridge National Laboratory has prepared a set of documents that evaluate a Partitioning-Transmutation (PT) fuel cycle relative to a Reference cycle employing conventional fuel-material recovery methods. The PT cycle uses enhanced recovery methods so that most of the long-lived actinides are recycled to nuclear power plants and transmuted to shorter-lived materials, thereby reducing the waste toxicity. This report compares the two fuel cycles on the basis of the short-term radiological and nonradiological risks they present to the public and to workers. The accidental radiological risk to the public is analyzed by estimating the probabilities of sets of accidents; the consequences are calculated using the CRAC code appropriately modified for the material composition. Routine radiological risks to the public are estimated from the calculated release amounts; the effects are calculated using the CRAC code. Radiological occupational risks are determined from prior experience, projected standards, and estimates of accident risk. Nonradiological risks are calculated from the number of personnel involved, historical experience, and epidemiological studies. The result of this analysis is that the short-term risk of PT is 2.9 times greater than that of the Reference cycle, primarily due to the larger amount of industry. This conclusion is strongly dominated by the nonradiological risk, which is about 150 times greater than the radiological risk. The absolute risk as estimated for the fuel cycle portions considered in this report is 0.91 fatalities/GWe-year for the PT cycle and 0.34 fatalities/GWe-year for the Reference cycle. This should be compared with Inhaber's estimate of 1.5 for nuclear and 150 for coal. All of the risks assumed here are associated with the production of one billion watts of electricity (GWe) per year.

Trends in state-level freight accident rates: An enhancement of risk factor development for RADTRAN

International Nuclear Information System (INIS)

Saricks, C.; Kvitek, T.

1991-01-01

Under the Nuclear Waste Policy Act, the Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) is concerned with understanding and managing risk as it applies to the shipment of spent commercial nuclear reactor fuel. Understanding risk in relation to mode and geography may provide opportunities to minimize radiological and non-radiological risks of transportation. To enhance such an understanding, a set of state-or waterway-specific accident, fatality, and injury rates (expressed as rates per shipment kilometer) by transportation mode and highway administrative class was developed, using publicly-available data bases. Adjustments made to accommodate miscoded or incomplete information in accident data are described, as well as the procedures for estimating state-level flow data. Results indicate that the shipping conditions under which spent fuel is likely to be transported should be less subject to accidents than the ''average'' shipment within mode. 10 refs., 3 tabs

Waste glass corrosion modeling: Comparison with experimental results

International Nuclear Information System (INIS)

Bourcier, W.L.

1993-11-01

A chemical model of glass corrosion will be used to predict the rates of release of radionuclides from borosilicate glass waste forms in high-level waste repositories. The model will be used both to calculate the rate of degradation of the glass, and also to predict the effects of chemical interactions between the glass and repository materials such as spent fuel, canister and container materials, backfill, cements, grouts, and others. Coupling between the degradation processes affecting all these materials is expected. Models for borosilicate glass dissolution must account for the processes of (1) kinetically-controlled network dissolution, (2) precipitation of secondary phases, (3) ion exchange, (4) rate-limiting diffusive transport of silica through a hydrous surface reaction layer, and (5) specific glass surface interactions with dissolved cations and anions. Current long-term corrosion models for borosilicate glass employ a rate equation consistent with transition state theory embodied in a geochemical reaction-path modeling program that calculates aqueous phase speciation and mineral precipitation/dissolution. These models are currently under development. Future experimental and modeling work to better quantify the rate-controlling processes and validate these models are necessary before the models can be used in repository performance assessment calculations

Waste management of ENM-containing solid waste in Europe

DEFF Research Database (Denmark)

Heggelund, Laura Roverskov; Boldrin, Alessio; Hansen, Steffen Foss

2015-01-01

the Danish nanoproduct inventory (www.nanodb.dk) to get a general understanding of the fate of ENM during waste management in the European context. This was done by: 1. assigning individual products to an appropriate waste material fraction, 2. identifying the ENM in each fraction, 3. comparing identified...... waste fractions with waste treatment statistics for Europe, and 4. illustrating the general distribution of ENM into incineration, recycling and landfilling. Our results indicate that ╲plastic from used product containers╡ is the most abundant and diverse waste fraction, comprising a variety of both...... nanoproducts and materials. While differences are seen between individual EU countries/regions according to the local waste management system, results show that all waste treatment options are significantly involved in nanowaste handling, suggesting that research activities should cover different areas...

Results of the freeze resistance test, swelling index and coefficient of permeability of finegrained mining waste reinforced with cements

Science.gov (United States)

Morman, Justyna

2018-04-01

The article presents the result of laboratory tests for mining waste with grain size of 0 to 2 mm stabilized with cement. Used for stabilization of cement CEM I 42.5 R and blast furnace cement CEM III / A 42.5N - LH / HSR / NA and a plasticizer sealant. Cement was added to the mining waste test in the proportions of 5 - 8% in relation to the skeleton's weight. For the cemented samples, the freeze resistance test, swelling index, coefficient of permeability and pH of water leachate were tested. The addition of a cement binder resulted in diminishing the water permeability of mining waste and limiting the leaching of fine particles from the material.

Results of geo-radio-monitoring for radioactive waste storage in large diameter boreholes in clayey ground

International Nuclear Information System (INIS)

Dmitriev, S.; Litinsky, Y.; Tkachenko, A.

2010-01-01

Document available in extended abstract form only. Full text of publication follows: The main purpose of the work carried out at the site of SUE MosSIA 'Radon' is to develop the system of geo-radio-monitoring for new type of storage facility (large diameter borehole) integrated into existing monitoring system of the whole site, check its effectiveness and improve the system, obtain initial results on safety aspects for using large diameter boreholes for RAW storage. Technology of large diameter boreholes (LDB) construction for low- and intermediate-level waste (LILW) isolation in moraine loams is being under development at SUE MosSIA 'Radon' site since the end of the last century. A project for construction of a demonstration unit for LILW storage in large diameter boreholes at the SUE MosSIA 'Radon' site in Sergiev Posad region has been developed taking into account specific site conditions. The main aim of the project is to develop the technology of LDB repository construction, operational procedures such as loading and retrieval, to develop and improve monitoring system for the new repository type, to get practical data on safety of radioactive wastes storage in new repositories, hermeticity of construction, and behavior of waste, waste packages, construction materials and near-field. In the case of LDB applications for LILW storage, the waste are removed from the scope of human activity into a stable geological medium. Waste are placed below the frost zone where damage of engineered barriers due to climatic factors is practically impossible. Two boreholes with 1.5 m internal diameter and 38 m depth have been drilled in 1997, equipped with engineering barriers including bentonite-concrete stone, licensed as storage facilities in 2003 and are in use now for solid and solidified RAW storage. Specific automated system of geo-radio-monitoring has been developed especially for the LDB-type repository, covering both the interior and the

Preliminary results from uranium/americium affinity studies under experimental conditions for cesium removal from NPP ''Kozloduy'' simulated wastes solutions

International Nuclear Information System (INIS)

Nikiforova, A.; Kinova, L.; Peneva, C.; Taskaeva, I.; Petrova, P.

2005-01-01

We use the approach described by Westinghouse Savannah River Company using ammonium molybdophosphate (AMP) to remove elevated concentrations of radioactive cesium to facilitate handling waste samples from NPP K ozloduy . Preliminary series of tests were carried out to determine the exact conditions for sufficient cesium removal from five simulated waste solutions with concentrations of compounds, whose complexing power complicates any subsequent processing. Simulated wastes solutions contain high concentrations of nitrates, borates, H 2 C 2 O 4 , ethylenediaminetetraacetate (EDTA) and Citric acid, according to the composition of the real waste from the NPP. On this basis a laboratory treatment protocol was created. This experiment is a preparation for the analysis of real waste samples. In this sense the results are preliminary. Unwanted removal of non-cesium radioactive species from simulated waste solutions was studied with gamma spectrometry with the aim to find a compromise between on the one hand the AMP effectiveness and on the other hand unwanted affinity to AMP of Uranium and Americium. Success for the treatment protocol is defined by proving minimal uptake of U and Am, while at the same time demonstrating good removal effectiveness through the use of AMP. Uptake of U and Am were determined as influenced by oxidizing agents at nitric acid concentrations, proposed by Savannah River National laboratory. It was found that AMP does not significantly remove U and Am when concentration of oxidizing agents is more than 0.1M for simulated waste solutions and for contact times inherent in laboratory treatment protocol. Uranium and Americium affinity under experimental conditions for cesium removal were evaluated from gamma spectrometric data. Results are given for the model experiment and an approach for the real waste analysis is chosen. Under our experimental conditions simulated wastes solutions showed minimal affinity to AMP when U and Am are most probably in

Results of Washington's phase two study on closure requirements for the Hanford commercial low-level waste facility

International Nuclear Information System (INIS)

Anderson, D.C.; Hana, S.L.

1989-01-01

This paper reports on the closure design objectives and cover alternatives resulting from the state of Washington's phase two study on closure and long-term care for the Hanford commercial low-level radioactive waste disposal facility. Four approaches to dealing with subsidence and two cover design alternatives are discussed in this paper, along with information on each layer of each cover. Objectives for closure of the Hanford low-level waste facility are also discussed

Technical solution for radioactive waste management resulting from the I-131 therapy

International Nuclear Information System (INIS)

Jerez V, P.; Lopez F, Y.; Quevedo G, J.; Betancourt H, L.

1996-01-01

The paper discusses a system designed for the collection and storage of biological wastes arising from the therapy with l-131. This system is based on the use of either retention or septic tanks, in which the waste is stored or delayed until the activity decays to acceptable levels, in order to comply with authorized limits established by the Regulatory Authority for discharge to environment. A method for estimating waste activity concentration as a function of the number of patients, the activity delivered to each one of them, as well as other parameter related to the system design are discussed. The general requirements to be met by the system are also included. (authors). 4 refs., 4 figs

Metals partitioning resulting from rotary kiln incineration of hazardous waste

International Nuclear Information System (INIS)

Richards, M.K.; Fournier, D.J. Jr.

1992-01-01

In response to the need for date on the partitioning of trace metals from hazardous waste incinerators, an extensive series of test was conducted in the summer of 1991 at the USEPA Incineration Research Facility (IRF) in Jefferson, Arkansas. These tests were conducted in the IRF's rotary kiln incinerator system (RKS) equipped with a pilot-scale Calvert Flux-Force/Condensation scrubber as the primary air pollution control system (APCS). The purpose of this test series was to extend the data base on trace metal partitioning and to investigate the effects of variations in incinerator operation on metal partitioning. Another objective was to evaluate the effectiveness of the scrubber for collecting flue gas metals. This series is a continuation of an ongoing IRF research program investigating trace metal partitioning and APCS collection efficiencies. Two previous test series were conducted using the RKS equipped with a venturi/packed-column scrubber and a single-state ionizing wet scrubber. The primary objective of this test series was to determine the fate of six hazardous and four nonhazardous trace metals fed to the RKS in a synthetic, organic-contaminated solid waste matrix. The six hazardous trace metals used were arsenic, barium, cadmium, chromium, mercury, and lead. The four nonhazardous trace metals--bismuth, copper, magnesium, and strontium--were included primarily to supply data to evaluate their potential for use as surrogates. The temperature, waste feed chlorine content, and scrubber pressure drop. The test program objectives were to identify. The partitioning of metals among kiln ash, scrubber liquor, and flue gas. Changes in metal partitioning related to variations in kiln exit gas temperature and waste feed chlorine content. The efficiency of the Calvert scrubber for collecting flue gas metals. The effects of scrubber pressure drop on metal collection efficiencies. 2 figs., 2 tabs

Importance of waste composition for Life Cycle Assessment of waste management solutions

DEFF Research Database (Denmark)

Bisinella, Valentina; Götze, Ramona; Conradsen, Knut

2017-01-01

The composition of waste materials has fundamental influence on environmental emissions associated with waste treatment, recycling and disposal, and may play an important role also for the Life Cycle Assessment (LCA) of waste management solutions. However, very few assessments include effects...... of the waste composition and waste LCAs often rely on poorly justified data from secondary sources. This study systematically quantifiesy the influence and uncertainty on LCA results associated with selection of waste composition data. Three archetypal waste management scenarios were modelled with the waste...... LCA model EASETECH based on detailed waste composition data from the literature. The influence from waste composition data on the LCA results was quantified with a step-wise Global Sensitivity Analysis (GSA) approach involving contribution, sensitivity, uncertainty and discernibility analyses...

Waste acceptance and waste loading for vitrified Oak Ridge tank waste

International Nuclear Information System (INIS)

Harbour, J.R.; Andrews, M.K.

1997-01-01

The Office of Science and Technology of the DOE has funded a joint project between the Oak Ridge National Laboratory (ORNL) and the Savannah River Technology Center (SRTC) to evaluate vitrification and grouting for the immobilization of sludge from ORNL tank farms. The radioactive waste is from the Gunite and Associated Tanks (GAAT), the Melton Valley Storage Tanks (MVST), the Bethel Valley Evaporator Service Tanks (BVEST), and the Old Hydrofractgure Tanks (OHF). Glass formulation development for sludge from these tanks is discussed in an accompanying article for this conference (Andrews and Workman). The sludges contain transuranic radionuclides at levels which will make the glass waste form (at reasonable waste loadings) TRU. Therefore, one of the objectives for this project was to ensure that the vitrified waste form could be disposed of at the Waste Isolation Pilot Plant (WIPP). In order to accomplish this, the waste form must meet the WIPP Waste Acceptance Criteria (WAC). An alternate pathway is to send the glass waste forms for disposal at the Nevada Test Site (NTS). A sludge waste loading in the feed of 6 wt percent will lead to a waste form which is non-TRU and could potentially be disposed of at NTS. The waste forms would then have to meet the requirements of the NTS WAC. This paper presents SRTC''s efforts at demonstrating that the glass waste form produced as a result of vitrification of ORNL sludge will meet all the criteria of the WIPP WAC or NTS WAC

Radiological dose assessment related to management of naturally occurring radioactive materials generated by the petroleum industry

International Nuclear Information System (INIS)

Smith, K.P.; Blunt, D.L.; Williams, G.P.

1996-09-01

A preliminary radiological dose assessment of equipment decontamination, subsurface disposal, landspreading, equipment smelting, and equipment burial was conducted to address concerns regarding the presence of naturally occurring radioactive materials (NORM) in production waste streams. The assessment estimated maximum individual dose equivalents for workers and the general public. Sensitivity analyses of certain input parameters also were conducted. On the basis of this assessment, it is concluded that (1) regulations requiring workers to wear respiratory protection during equipment cleaning operations are likely to result in lower worker doses, (2) underground injection and downhole encapsulation of NORM wastes present a negligible risk to the general public, and (3) potential doses to workers and the general public related to smelting NORM-contaminated equipment can be controlled by limiting the contamination level of the initial feed. It is recommended that (1) NORM wastes be further characterized to improve studies of potential radiological doses; (2) states be encouraged to permit subsurface disposal of NORM more readily, provided further assessments support this study; results; (3) further assessment of landspreading NORM wastes be conducted; and (4) the political, economic, sociological, and nonradiological issues related to smelting NORM-contaminated equipment be studied to fully examine the feasibility of this disposal option

Management of the radioactive waste resulting from the Romanian VVR-S research reactor decommissioning

International Nuclear Information System (INIS)

Ene, D.; Cepraga, D.G.

2002-01-01

The paper consists in a waste study of the Romanian VVR-S reactor which will be prepared for decommissioning operations after the permanent shutdown (23.12.1997). Calculations were carried out to determine the activity arising from neutron activation of structural materials inside the reactor, considering the design of the facility and its operating rules. To this end, the following method was used: i) Neutron flux distribution within the reactor was calculated using the DORT transport code, based on DLC23 shielding library relating to three cylindrical reference systems of the reactor structure: reactor core, horizontal tube and thermal column; ii) Calculation of the activity of each reactor component at different cooling times was performed by the ANITA2000 code, using the neutron flux, compositional data for each material and the power history of the reactor; iii) Unconditional clearance indexes for all material at various cooling times were calculated using the clearance levels defined in IAEA-TECDOC-855; iv) Total activities and masses by material type, within the waste category and for each decay time were calculated by summation of the data previously classified for each reactor component. The resulting activation inventory and waste masses, falling in IAEA defined waste categories are presented in the paper at periods of 100 days, and 6, 10, 25, and 50 years after reactor the shutdown. For some components of the reactor as: aluminum central vessel, the central iron shielding ring, the time behaviour of both the fin spatial activity distribution and the radionuclide contributions to the total activity are plotted in the paper. (author)

Results of technical and economical examinations for substantiation of special plant design for reprocessing and radioactive wastes disposal

International Nuclear Information System (INIS)

Galkin, A.V.; Baldov, A.N.

2001-01-01

In the paper the results of technical and economical examinations for substantiation of special plant design for reprocessing and radioactive wastes disposal are presented. Ground for the examination conducting was Health of Nation Programme ratified by the President and a number of Governmental decisions. The special plant is planned in the Mangystau Region. In the framework of feasibility study the data base by the worldwide known technologies was implemented, on reprocessing and experience of radioactive waste disposal. The technical requirements for the special plant construction are determined. The alternative options by structure content and site location of the special plant and radioactive waste disposal are cited

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

International Nuclear Information System (INIS)

Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian

2016-01-01

This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na_2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl- in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease the waste

Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

Energy Technology Data Exchange (ETDEWEB)

Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-01

This report describes the scientific basis underlying the approach being followed to design and develop “advanced” glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na₂O content than glass compositions used previously to provide enough Na+ to react with all of the Cl– in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease

Influence of assumptions about household waste composition in waste management LCAs

International Nuclear Information System (INIS)

Slagstad, Helene; Brattebø, Helge

2013-01-01

Highlights: ► Uncertainty in waste composition of household waste. ► Systematically changed waste composition in a constructed waste management system. ► Waste composition important for the results of accounting LCA. ► Robust results for comparative LCA. - Abstract: This article takes a detailed look at an uncertainty factor in waste management LCA that has not been widely discussed previously, namely the uncertainty in waste composition. Waste composition is influenced by many factors; it can vary from year to year, seasonally, and with location, for example. The data publicly available at a municipal level can be highly aggregated and sometimes incomplete, and performing composition analysis is technically challenging. Uncertainty is therefore always present in waste composition. This article performs uncertainty analysis on a systematically modified waste composition using a constructed waste management system. In addition the environmental impacts of several waste management strategies are compared when applied to five different cities. We thus discuss the effect of uncertainty in both accounting LCA and comparative LCA. We found the waste composition to be important for the total environmental impact of the system, especially for the global warming, nutrient enrichment and human toxicity via water impact categories.

Drug waste minimisation and cost-containment in Medical Oncology: Two-year results of a feasibility study

Directory of Open Access Journals (Sweden)

Mansutti Mauro

2008-04-01

Full Text Available Abstract Background Cost-containment strategies are required to face the challenge of rising drug expenditures in Oncology. Drug wastage leads to economic loss, but little is known about the size of the problem in this field. Methods Starting January 2005 we introduced a day-to-day monitoring of drug wastage and an accurate assessment of its costs. An internal protocol for waste minimisation was developed, consisting of four corrective measures: 1. A rational, per pathology distribution of chemotherapy sessions over the week. 2. The use of multi-dose vials. 3. A reasonable rounding of drug dosages. 4. The selection of the most convenient vial size, depending on drug unit pricing. Results Baseline analysis focused on 29 drugs over one year. Considering their unit price and waste amount, a major impact on expense was found to be attributable to six drugs: cetuximab, docetaxel, gemcitabine, oxaliplatin, pemetrexed and trastuzumab. The economic loss due to their waste equaled 4.8% of the annual drug expenditure. After the study protocol was started, the expense due to unused drugs showed a meaningful 45% reduction throughout 2006. Conclusion Our experience confirms the economic relevance of waste minimisation and may represent a feasible model in addressing this issue. A centralised unit of drug processing, the availability of a computerised physician order entry system and an active involvement of the staff play a key role in allowing waste reduction and a consequent, substantial cost-saving.

Results of detailed ground geophysical surveys for locating and differentiating waste structures in waste management area 'A' at Chalk River Laboratories, Ontario

International Nuclear Information System (INIS)

Tomsons, D.K.; Street, P.J.; Lodha, G.S.

1999-01-01

Waste Management Area 'A' (WMA 'A'), located in the outer area of the Chalk River Laboratories (CRL) was in use as a waste burial site from 1946 to 1955. Waste management structures include debris-filled trenches, concrete bunkers and miscellaneous contaminated solid materials, and ditches and pits used for liquid dispersal. In order to update historical records, it was proposed to conduct detailed ground geophysical surveys to define the locations of waste management structures in WMA 'A', assist in planning of the drilling and sampling program to provide ground truth for the geophysics investigation and to predict the nature and locations of unknown/undefined shallow structures. A detailed ground geophysical survey grid was established with a total of 127 grid lines, oriented NNE and spaced one metre apart. The geophysical surveys were carried out during August and September, 1996. The combination of geophysical tools used included the Geonics EM61 metal detector, the GSM-19 magnetometer/gradiometer and a RAMAC high frequency ground penetrating radar system. The geophysical surveys were successful in identifying waste management structures and in characterizing to some extent, the composition of the waste. The geophysical surveys are able to determine the presence of most of the known waste management structures, especially in the western and central portions of the grid which contain the majority of the metallic waste. The eastern portion of the grid has a completely different geophysical character. While historical records show that trenches were dug, they are far less evident in the geophysical record. There is clear evidence for a trench running between lines 30E and 63E at 70 m. There are indications from the radar survey of other trench-like structures in the eastern portion. EM61 data clearly show that there is far less metallic debris in the eastern portion. The geophysical surveys were also successful in identifying previously unknown locations of waste

Fertilization effects of organic waste resources and bottom wood ash: results from a pot experiment

Directory of Open Access Journals (Sweden)

Eva Brod

2012-12-01

Full Text Available We conducted a pot experiment to study the fertilization effects of four N- and P-rich organic waste resources alone and in combination with K-rich bottom wood ash at two application rates (150 kg N ha–1 + 120 kg K ha–1, 300 kg N ha-1 + 240 kg K ha–1. Plant-available N was the growth-limiting factor. 48–73% of N applied with meat and bone meal (MBM and composted fish sludge (CFS was taken up in aboveground biomass, resulting in mineral fertilizer equivalents (MFE% of 53–81% for N uptake and 61–104% for yield. MFE% of MBM and CFS decreased for increasing application rates. Two industrial composts had weak N fertilization effects and are to be considered soil conditioners rather than fertilizers. Possible P and K fertilization effects of waste resources were masked by the soil’s ability to supply plant-available P and K, but effects on plant-available P and K contents in soil suggest that the waste resources may have positive effects under more nutrient-deficient conditions.

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

Energy Technology Data Exchange (ETDEWEB)

Rechard, R.P. [ed.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

Results of questionaire survey for the measurement of radioactivity in waste water

International Nuclear Information System (INIS)

1992-01-01

A questionaire for radioactivity in waste water was sent to 388 facilities, including 158 medical facilities, and all (100%) answered. Information requested included: (1) kinds and annual usage of unsealed RI, (2) measuring method of radioactivity in waste water, (3) kinds of measuring instruments and the detection limits, (4) prior treatment of measurement materials, (5) level of radioactive waste exhausted during 3 months, (6) personnel and time per month required for radioactivity measurement, (7) problems and comments in waste water management, and (8) kinds of facilities. A total of 36 unsealed RI were used. The most commonly used RI was I-125 (n=240), followed by H-3 (n=189) and P-32 (n=179). Annual level of RI was 4 GBq or less in 90% of the facilities. The most common method for measuring radioactivity was sampling method (n=241). The most common instrument for measuring radioactivity was a gamma counter for I-125 (45% of the facilities), and a liquid scintillation counter for P-32 (80%) and for C-14 and H-3 (90%). The detection limits for I-125 exceeded the radioactivity limits in 24% of the facilities. The amount of sampler was 5 cc or less in 80% of the facilities. Prio treatment was not carried out in 62.7%. Prior treatment methods reported were enrichment, evaporation, pH adjustment, and sedimentation. Half of the facilities exhausted 10 cm 3 or less of waste water during 3 months. The number of persons engaging in radioactivity measurement per month was reported to be one in 282 facilities (87%). (N.K.)

Waste management, waste resource facilities and waste conversion processes

International Nuclear Information System (INIS)

Demirbas, Ayhan

2011-01-01

In this study, waste management concept, waste management system, biomass and bio-waste resources, waste classification, and waste management methods have been reviewed. Waste management is the collection, transport, processing, recycling or disposal, and monitoring of waste materials. A typical waste management system comprises collection, transportation, pre-treatment, processing, and final abatement of residues. The waste management system consists of the whole set of activities related to handling, treating, disposing or recycling the waste materials. General classification of wastes is difficult. Some of the most common sources of wastes are as follows: domestic wastes, commercial wastes, ashes, animal wastes, biomedical wastes, construction wastes, industrial solid wastes, sewer, biodegradable wastes, non-biodegradable wastes, and hazardous wastes.

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

Energy Technology Data Exchange (ETDEWEB)

Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

2014-01-13

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

Methodology for Safety Assessment Applied to Predisposal Waste Management. Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) 2004–2010)

International Nuclear Information System (INIS)

2015-12-01

Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) (2004–2010) The IAEA’s progamme on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) focused on approaches and mechanisms for application of safety assessment methodologies for the predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts, which have since been incorporated into IAEA Safety Standards Series No. GSG-3, Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste. In 2005, an initial specification was developed for the Safety Assessment Framework (SAFRAN) software tool to apply the SADRWMS flowcharts. In 2008, an in-depth application of the SAFRAN tool and the SADRWMS methodology was carried out on the predisposal management facilities of the Thailand Institute of Nuclear Technology Radioactive Waste Management Centre (TINT Facility). This publication summarizes the content and outcomes of the SADRWMS programme. The Chairman’s Report of the SADRWMS Project and the Report of the TINT test case are provided on the CD-ROM which accompanies this report

Waste Management's LNG Truck Fleet: Final Results

Energy Technology Data Exchange (ETDEWEB)

Chandler, K. [Battelle (US); Norton, P. [National Renewable Energy Laboratory (US); Clark, N. [West Virginia University (US)

2001-01-25

Waste Management, Inc., began operating a fleet of heavy-duty LNG refuse trucks at its Washington, Pennsylvania, facility. The objective of the project was to provide transportation professionals with quantitative, unbiased information on the cost, maintenance, operational, and emissions characteristics of LNG as one alternative to conventional diesel for heavy-duty trucking applications.

Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

International Nuclear Information System (INIS)

Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas; Jegou, Christophe

2007-01-01

Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF 3 medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

Study of the use waste resulting from the mining of emerald for the production refractory ceramic

International Nuclear Information System (INIS)

Esteves, P.J.C.; Coelho, R.E.; Cruz, R.M.S.; Cavalcanti, R.F.

2009-01-01

Full text: The great impact caused by excess mineral waste in ambient of the emerald exploration, in determined locals of Brazil, where are deposited, it has caused inconvenience to their various people residents. The jungles, rivers and lakes are directly harmed by the aggressions imposed by neglect in the destination of such waste. Considering the importance of the issue outlined to the goal of this work, this paper can back report a study for utilizing emerald waste, focused the possibility manufacture for obtained refractory ceramic. The results show that the specimens prepared by the ball milling, cold pressing and sintering method had better high temperature properties, due to a higher mica volume percent and finer crystallite size. Specimens it was characterized by X-ray diffractometer and fluorescence. Test was realized in the materials, submitted in high temperature was observed good thermal stability, the processed ceramics could be recommended for the adequate applications. (author)

Disposal of radwastes and recycling of wastes and structural materials -fundamental principles, concepts, results

International Nuclear Information System (INIS)

Schaller, G.; Arens, G.; Brennecke, P.; Goertz, R.; Poschner, J.; Thieme, M.

1997-01-01

This report describes the German concept for the disposal of radioactive waste, and the re-use or recycling of contaminated materials. All radioactive waste can be disposed of in deep geological formations (practised at ERAM disposal site, planned for Konrad disposal site). Radioactively contaminated material below clearance levels can proceed for disposal at waste disposal sites and incineration plants, or for re-use and recycling, especially where the material consists of contaminated steel and of buildings. The basic principles (dose limits and model structures for deriving recommendations), reference values, or limits are described. The latest concepts are described in greater detail. Waste management in Germany is compared with international concepts. (orig.) [de

Finite-element model evaluation of barrier configurations to reduce infiltration into waste-disposal structures: preliminary results and design considerations

International Nuclear Information System (INIS)

Lu, A.H.; Phillips, S.J.; Adams, M.R.

1982-09-01

Barriers to reduce infiltration into waste burial disposal structures (trenches, pits, etc.) may be required to provide adequate waste confinement. The preliminary engineering design of these barriers should consider interrelated barrier performance factors. This paper summarizes preliminary computer simulation activities to further engineering barrier design efforts. Several barrier configurations were conceived and evaluated. Models were simulated for each barrier configuration using a finite element computer code. Results of this preliminary evaluation indicate that barrier configurations, depending on their morphology and materials, may significantly influence infiltration, flux, drainage, and storage of water through and within waste disposal structures. 9 figures

Product consistency test and toxicity characteristic leaching procedure results of the ceramic waste form from the electrometallurgical treatment process for spent fuel

International Nuclear Information System (INIS)

Johnson, S. G.; Adamic, M. L.: DiSanto, T.; Warren, A. R.; Cummings, D. G.; Foulkrod, L.; Goff, K. M.

1999-01-01

The ceramic waste form produced from the electrometallurgical treatment of sodium bonded spent fuel from the Experimental Breeder Reactor-II was tested using two immersion tests with separate and distinct purposes. The product consistency test is used to assess the consistency of the waste forms produced and thus is an indicator of a well-controlled process. The toxicity characteristic leaching procedure is used to determine whether a substance is to be considered hazardous by the Environmental Protection Agency. The proposed high level waste repository will not be licensed to receive hazardous waste, thus any waste forms destined to be placed there cannot be of a hazardous nature as defined by the Resource Conservation and Recovery Act. Results are presented from the first four fully radioactive ceramic waste forms produced and from seven ceramic waste forms produced from cold surrogate materials. The fully radioactive waste forms are approximately 2 kg in weight and were produced with salt used to treat 100 driver subassemblies of spent fuel

Results of sampling the contents of the liquid low-level waste evaporator feed tank W-22 at ORNL

International Nuclear Information System (INIS)

Sears, M.B.

1996-09-01

This report summarizes the results of the fall 1994 sampling of the contents of the liquid low- level waste (LLLW) tank W-22 at the Oak Ridge National Laboratory (ORNL). Tank W-22 is the central collection and holding tank for LLLW at ORNL before the waste is transferred to the evaporators. Samples of the tank liquid and sludge were analyzed to determine (1) the major chemical constituents, (2) the principal radionuclides, (3) the metals listed on the U.S. Environmental Protection Agency (EPA) Contract Laboratory Program Inorganic Target Analyte List, (4) organic compounds, and (5) some physical properties. The organic chemical characterization consisted of the determinations of the EPA Contract Laboratory Program Target Compound List semivolatile compounds, pesticides, and polychlorinated biphenyls (PCBs). Water-soluble volatile organic compounds were also determined. Information provided in this report forms part of the technical basis in support of (1) waste management for the active LLLW system and (2) planning for the treatment and disposal of the waste

Independent monitoring of a release from the waste isolation pilot plant in New Mexico, USA. Results and purpose

Energy Technology Data Exchange (ETDEWEB)

Thakur, Punam; Ballard, Sally [Carlsbad Environmental Monitoring and Research Center, Carlsbad, NM (United States)

2015-07-01

The Waste Isolation Pilot Plant (WIPP) is a transuranic (TRU) waste repository operated by the U.S. Department of Energy (DOE). The repository is emplacing defense-related transuranic (TRU) wastes into a bedded salt formation approximately 655 m (2150 ft.) below the surface of the Earth. Located near Carlsbad, New Mexico, an area with less than 30,000 people, the WIPP facility is licensed to accept TRU waste with activity concentrations of alpha-emitting isotopes >3700 Bq/m{sup 3} (> 100 nCi/g) and half-life >20 years. The upper waste acceptance limit is 0.85 TBq/liter (<23 Ci/liter) of total activity and 10 Sv/hr dose rate on contact. The repository, which opened in March 1999 will eventually contain the equivalent of ∝176,000 m{sup 3} of TRU waste. The vast majority of the waste disposed in the WIPP repository is ''contact-handled'' waste, meaning it has a surface dose rate less than 2 mSv per hour. Local acceptance of WIPP is in part due to an independent environmental monitoring program that began before and continues after WIPP began receiving nuclear waste. This independent monitoring is being conducted by the Carlsbad Environmental Monitoring and Research Center (CEMRC), which is associated with New Mexico State University. CEMRC is funded by DOE through a grant process that respects its independence in carrying out and reporting the results of environmental monitoring at and near the WIPP site. The primary focus of CEMRC monitoring is on airborne radioactive particulate; however other pathways are also monitored. Pre-disposal baseline data of various anthropogenic radionuclides present in the WIPP environment is essential for the proper evaluation of the WIPP integrity. These data are compared against disposal phase data to assess whether or not there is any radiological impact from the presence of WIPP on workers and on the regional public. The program has capabilities to detect radionuclides rapidly in case of accidental releases

An evaluation on the disposal alternatives for low- and intermediate- level radwaste (II)

International Nuclear Information System (INIS)

Park, Hun Hwee; Han, Kyung Won; Hahn, Pil Soo; Lee, Han Soo; Cho, Won Jin; Lee, Jae Dwan; Park, Chung Kyun; Lee, Myung Joo; Choi, Heui Joo; Lee, Youn Myoung

1988-02-01

An evaluation on the radioactive waste disposal alternatives for the low-and intermediate level wastes being produced from nuclear power generation and radioisotope application was carried out in view of the radiological safety, socio-political aspects and repository construction economics. Three types of possible alternatives-sample shallow land disposal method, engineered shallow land disposal method and engineered rock cavern disposal method are investigated. The safety assessment consists of radiological dose calculation and nonradiological impacts which is expressed as total number of injuries and fatalities during construction, operation and transportation. The sociopolitical assessment is done in terms of site conditions including easiness for land acquisition, technical feasibility and public acceptance. The economic assessment is performed by cost comparison regarding land acquisition, construction, operation and closure for each alternatives. The evaluation shows that engineered rock cavern disposal method has remarkable favour in safety than others. And also an integrated evaluation using AHP results the engineered rock cavern disposal method as the most favorable option

Genotoxicity studies in semiconductor industry. 1. In vitro mutagenicity and genotoxicity studies of waste samples resulting from plasma etching

Energy Technology Data Exchange (ETDEWEB)

Braun, R.; Huettner, E.M.; Merten, H.; Raabe, F. (Institute of Plant Genetics and Crop Plant Research, Gatersleben (Germany))

1993-07-01

Solid waste samples taken from the etching reactor, the turbo pump, and the waste air system of a plasma etching technology line in semiconductor production were studied as to their genotoxic properties in a bacterial repair test, in the Ames/Salmonella microsome assay, in the SOS chromotest, in primary mouse hepatocytes, and in Chinese hamster V79 cell cultures. All three waste samples were found to be active by inducing of unscheduled DNA-synthesis in mouse hepatocytes in vitro. In the bacterial rec-type repair test with Proteus mirabilis, waste samples taken from the turbo pump and the vacuum pipe system were not genotoxic. The waste sample taken from the chlorine-mediated plasma reactor was clearly positive in the bacterial repair assay and in the SOS chromotest with Escherichia coli. Mutagenic activity was demonstrated for all samples in the presence and absence of S9 mix made from mouse liver homogenate. Again, highest mutagenic activity was recorded for the waste sample taken from the plasma reactor, while samples collected from the turbo pump and from the waste air system before dilution and liberation of the air were less mutagenic. For all samples chromosomal damage in V79 cells was not detected, indicating absence of clastogenic activity in vitro. Altogether, these results indicate generation of genotoxic and mutagenic products as a consequence of chlorine-mediated plasma etching in the microelectronics industry and the presence of genotoxins even in places distant from the plasma reactor. Occupational exposure can be expected both from the precipitated wastes and from chemicals reaching the environment with the air stream.

Calculation of projected waste loads for transuranic waste management alternatives

International Nuclear Information System (INIS)

Hong, K.; Kotek, T.; Koebnick, B.; Wang, Y.; Kaicher, C.

1995-01-01

The level of treatment and the treatment and interim storage site configurations (decentralized, regional, or centralized) impact transuranic (TRU) waste loads at and en route to sites in the US Department of Energy (DOE) complex. Other elements that impact waste loads are the volume and characteristics of the waste and the unit operation parameters of the technologies used to treat it. Projected annual complexwide TRU waste loads under various TRU waste management alternatives were calculated using the WASTEunderscoreMGMT computational model. WASTEunderscoreMGMT accepts as input three types of data: (1) the waste stream inventory volume, mass, and contaminant characteristics by generating site and waste stream category; (2) unit operation parameters of treatment technologies; and (3) waste management alternative definitions. Results indicate that the designed capacity of the Waste Isolation Pilot Plant, identified under all waste management alternatives as the permanent disposal facility for DOE-generated TRU waste, is sufficient for the projected complexwide TRU waste load under any of the alternatives

Economic analysis of waste management alternatives for reprocessing wastes

International Nuclear Information System (INIS)

McKee, R.W.; Clark, L.L.; Daling, P.M.; Nesbitt, J.F.; Swanson, J.L.

1984-02-01

This study describes the results of a cost analysis of a broad range of alternatives for management of reprocessing wastes that would require geologic repository disposal. The intent was to identify cost-effective alternatives and the costs of potential repository performance requirements. Four integrated treatment facility alternatives for transuranic (TRU) wastes are described and compared. These include no treatment, compaction, incineration, and hulls melting. The advantages of reducing high-level wastes (HLW) volume are also evaluated as are waste transportation alternatives and several performance-related alternatives for emplacing waste in a basalt repository. Results show (1) that system costs for disposal of reprocessing waste are likely to be higher than those for disposal of spent fuel; (2) that volume reduction is cost-effective for both remote-handled (RH) TRU wastes and HLW, and that rail transport for HLW is more cost-effective than truck transport; (3) that coemplacement of RH-TRU wastes with HLW does not have a large cost advantage in a basalt repository; and (4) that, relative to performance requirements, the cost impact for elimination of combustibles is about 5%, long-lived containers for RH-TRU wastes can increase repository costs 10% to 20%, and immediate backfill compared to delayed backfill (bentonite/basalt) around the HLW canisters would increase repository costs up to 10% or overall system costs up to about 5%. 13 references, 4 figures, 12 tables

Nuclear waste and hazardous waste in the public perception

International Nuclear Information System (INIS)

Kruetli, Pius; Seidl, Roman; Stauffacher, Michael

2015-01-01

The disposal of nuclear waste has gained attention of the public for decades. Accordingly, nuclear waste has been a prominent issue in natural, engineer and social science for many years. Although bearing risks for todays and future generations hazardous waste in contrast is much less an issue of public concern. In 2011, we conducted a postal survey among Swiss Germans (N = 3.082) to learn more about, how nuclear waste is perceived against hazardous waste. We created a questionnaire with two versions, nuclear waste and hazardous waste, respectively. Each version included an identical part with well-known explanatory factors for risk perception on each of the waste types separately and additional questions directly comparing the two waste types. Results show that basically both waste types are perceived similarly in terms of risk/benefit, emotion, trust, knowledge and responsibility. However, in the direct comparison of the two waste types a complete different pattern can be observed: Respondents perceive nuclear waste as more long-living, more dangerous, less controllable and it, furthermore, creates more negative emotions. On the other hand, respondents feel more responsible for hazardous waste and indicate to have more knowledge about this waste type. Moreover, nuclear waste is perceived as more carefully managed. We conclude that mechanisms driving risk perception are similar for both waste types but an overarching negative image of nuclear waste prevails. We propose that hazardous waste should be given more attention in the public as well as in science which may have implications on further management strategies of hazardous waste.

Nuclear waste and hazardous waste in the public perception

Energy Technology Data Exchange (ETDEWEB)

Kruetli, Pius; Seidl, Roman; Stauffacher, Michael [ETH Zurich (Switzerland). Inst. for Environmental Decisions

2015-07-01

The disposal of nuclear waste has gained attention of the public for decades. Accordingly, nuclear waste has been a prominent issue in natural, engineer and social science for many years. Although bearing risks for todays and future generations hazardous waste in contrast is much less an issue of public concern. In 2011, we conducted a postal survey among Swiss Germans (N = 3.082) to learn more about, how nuclear waste is perceived against hazardous waste. We created a questionnaire with two versions, nuclear waste and hazardous waste, respectively. Each version included an identical part with well-known explanatory factors for risk perception on each of the waste types separately and additional questions directly comparing the two waste types. Results show that basically both waste types are perceived similarly in terms of risk/benefit, emotion, trust, knowledge and responsibility. However, in the direct comparison of the two waste types a complete different pattern can be observed: Respondents perceive nuclear waste as more long-living, more dangerous, less controllable and it, furthermore, creates more negative emotions. On the other hand, respondents feel more responsible for hazardous waste and indicate to have more knowledge about this waste type. Moreover, nuclear waste is perceived as more carefully managed. We conclude that mechanisms driving risk perception are similar for both waste types but an overarching negative image of nuclear waste prevails. We propose that hazardous waste should be given more attention in the public as well as in science which may have implications on further management strategies of hazardous waste.

Nevada Test Site Environmental Report 2007

Energy Technology Data Exchange (ETDEWEB)

Cathy Wills

2008-09-01

The Nevada Test Site Environmental Report 2007 (NTSER) was prepared to meet the information needs of the public and the requirements and guidelines of the U.S. Department of Energy (DOE) for annual site environmental reports. It was prepared by National Security Technologies, LLC (NSTec). This Executive Summary presents the purpose of the document, the major programs conducted at the Nevada Test Site (NTS), NTS key environmental initiatives, radiological releases and potential doses to the public resulting from site operations, a summary of nonradiological releases, implementation status of the NTS Environmental Management System, a summary of compliance with environmental regulations, pollution prevention and waste minimization accomplishments, and significant environmental accomplishments. Much of the content of this Executive Summary is also presented in a separate stand-alone pamphlet titled Nevada Test Site Environmental Report Summary 2007. This NTSER was prepared to satisfy DOE Order 231.1A, Environment, Safety and Health Reporting. Its purpose is to (1) report compliance status with environmental standards and requirements, (2) present results of environmental monitoring of radiological and nonradiological effluents, (3) report estimated radiological doses to the public from releases of radioactive material, (4) summarize environmental incidents of noncompliance and actions taken in response to them, (5) describe the NTS Environmental Management System and characterize its performance, and (6) highlight significant environmental programs and efforts. This report meets these objectives for the NTS and three offsite Nevada facilities mentioned in this report.

Nevada Test Site Environmental Report 2007

International Nuclear Information System (INIS)

Cathy Wills

2008-01-01

The Nevada Test Site Environmental Report 2007 (NTSER) was prepared to meet the information needs of the public and the requirements and guidelines of the U.S. Department of Energy (DOE) for annual site environmental reports. It was prepared by National Security Technologies, LLC (NSTec). This Executive Summary presents the purpose of the document, the major programs conducted at the Nevada Test Site (NTS), NTS key environmental initiatives, radiological releases and potential doses to the public resulting from site operations, a summary of nonradiological releases, implementation status of the NTS Environmental Management System, a summary of compliance with environmental regulations, pollution prevention and waste minimization accomplishments, and significant environmental accomplishments. Much of the content of this Executive Summary is also presented in a separate stand-alone pamphlet titled Nevada Test Site Environmental Report Summary 2007. This NTSER was prepared to satisfy DOE Order 231.1A, Environment, Safety and Health Reporting. Its purpose is to (1) report compliance status with environmental standards and requirements, (2) present results of environmental monitoring of radiological and nonradiological effluents, (3) report estimated radiological doses to the public from releases of radioactive material, (4) summarize environmental incidents of noncompliance and actions taken in response to them, (5) describe the NTS Environmental Management System and characterize its performance, and (6) highlight significant environmental programs and efforts. This report meets these objectives for the NTS and three offsite Nevada facilities mentioned in this report

Overview of french P and T programme and results for waste management

International Nuclear Information System (INIS)

Warin, D.; Courtois, C.

2005-01-01

We will present here the French program and an update on the progress made by the research conducted on partitioning and transmutation. Studies on partitioning and transmutation aim at isolating the most radio toxic long-lived elements present in the waste then at transmuting them through recycling in nuclear reactors, in order to change them into non-radioactive or shorter-lived elements. The partitioning of minor actinides (americium, curium and neptunium), followed by their transmutation, would reduce to a few hundred years the time necessary for the radiotoxicity of the vitrified waste to become similar to that contained in the natural uranium ore originally used. The feasibility of partitioning, which did not appear easily accessible at the time the research began since lanthanides and actinides have rather similar chemical properties, was nevertheless demonstrated in 2001 thanks to a series of tests conducted on solutions of dissolved spent fuel, in the CEA Atalante facility at Marcoule. The 2002-2005 program encompasses technological demonstration of the selected liquid-liquid process, with representative equipment, and economic evaluation of industrial implementation of partitioning. Studies on transmutation, which were initiated before the 1991 Law, rapidly led to concluding that transmutation of minor actinides (Americium, Curium, and Neptunium) was feasible in particular in fast neutron spectra. Results obtained confirm that the feasibility of transmutation is demonstrated, both in pressurized-water reactors (recycling and transmutation of plutonium, optionally but with more difficulty of americium and neptunium) and in advanced systems of nuclear-energy production (GEN IV fast-spectrum reactors, with recycling and transmutation of all heavy nuclides, uranium, plutonium, the minor actinides) or in dedicated incinerator reactors, either critical or sub critical. Work on transmutation is now focusing on technical elements necessary for the demonstration of

Effluent Management Facility Evaporator Bottom-Waste Streams Formulation and Waste Form Qualification Testing

Energy Technology Data Exchange (ETDEWEB)

Saslow, Sarah A.; Um, Wooyong; Russell, Renee L.

2017-08-02

This report describes the results from grout formulation and cementitious waste form qualification testing performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates three grout formulations proposed for wide-range treatment of different waste stream compositions expected for the Hanford Effluent Management Facility (EMF) evaporator bottom waste. This work supports the technical development need for alternative disposition paths for the EMF evaporator bottom wastes and future direct feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious waste form qualification testing. The work contained within this report relates to waste form development and testing, and does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY 2017 and future waste form development efforts. The provided results and data should be used by (1) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (2) decision makers interested in off-site waste form disposal, and (3) the U.S. Department of Energy, their Hanford Site contractors and stakeholders as they assess the IDF PA program at the Hanford Site. The results reported help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottom waste, and improve the technical defensibility of long-term waste form risk estimates.

The study of the container types used for transport and final disposal of the radioactive wastes resulting from decommissioning of nuclear facilities

International Nuclear Information System (INIS)

Postelnicu, C.

1998-01-01

The purpose of the present paper is to select from a variety of package forms and capacities some containers which will be used for transport and disposal of the radioactive wastes resulting from decommissioning of nuclear facilities into the National Repository for Radioactive Waste - Baita, Bihor county. Taken into account the possibilities of railway and / or road transport and waste disposal in our country, detailed container classification was given in order to use them for radioactive waste transport and final disposal from decommissioning of IFIN-HH Research Reactor. (author)

Supplemental results of the human health risk analysis for the U.S. Department of Energy draft waste management programmatic environmental impact statement

International Nuclear Information System (INIS)

1995-08-01

This report is intended as an information supplement to the human health risk analysis performed for the US Department of Energy's Draft Waste Management Programmatic Environmental Impact Statement for Managing Treatment, Storage, and Disposal of Radioactive and Hazardous Waste, hereinafter called the PEIS. This report provides the installation-by-installation human health risk analysis results from which the risk estimate summaries for the PEIS were drawn. Readers should bear in mind that the risk estimates presented here are the result of a program-wide (as opposed to site-specific) study. They are based on best available data; systematically applied assumptions; and professional judgment about DOE waste inventories, waste volumes generated annually, currently available treatment and disposal technologies, technical limitations of treatment, and facility capacities across the numerous installations in the DOE complex

PROFILE OF PLASTIC WATER BOTTLES WASTES PROCESSING BUSINESS UNIT FOR WASTE PICKERS

Directory of Open Access Journals (Sweden)

Herijanto P.

2017-09-01

Full Text Available Used plastic water bottles waste pickers can be categorized as one of the informal sector’s component. They work for themselves by picking up used water bottles and selling them to the waste collectors. The problem to be solved in this research is How the Most Appropriate Used Plastic Water Bottles Business Model for Waste Pickers Is that enables them to be categorized as formal sector. From the result of the interview with 120 waste pickers, 96 results were qualified to be analyzed. The interview was located in several waste collectors, which were visited by waste pickers at certain hours. The data were analyzed descriptively based on six business aspects. Specifically for production facilities, Quality Function Deployment (QFD and Value Engineering (VE analysis were performed. The results of the analysis indicate that the business is practicable for waste pickers and has the potential to enable them run a formal business sector.

Improved Process Used to Treat Aqueous Mixed Waste Results in Cost Savings and Improved Worker Safety

International Nuclear Information System (INIS)

Hodge, D.S.; Preuss, D.E.; Belcher, K.J.; Rock, C.M.; Bray, W.S.; Herman, J.P.

2006-01-01

This paper describes an improved process implemented at Argonne National Laboratory (ANL) to treat aqueous mixed waste. This waste is comprised of radioactively-contaminated corrosive liquids with heavy metals. The Aqueous Mixed Waste Treatment System (AMWTS) system components include a reaction tank and a post-treatment holding tank with ancillary piping and pumps; and a control panel with pumping/mixing controls; tank level, temperature and pH/Oxidation Reduction Potential (ORP) indicators. The process includes a neutralization step to remove the corrosive characteristic, a chromium reduction step to reduce hexavalent chromium to trivalent chromium, and a precipitation step to convert the toxic metals into an insoluble form. Once the toxic metals have precipitated, the resultant sludge is amenable to stabilization and can be reclassified as a low-level waste if the quantity of leachable toxic metals, as determined by the TCLP, is below Universal Treatment Standards (UTS). To date, six batches in eight have passed the UTS. The AMWTS is RCRA permitted and allows for the compliant treatment of mixed waste prior to final disposal at a Department of Energy (DOE) or commercial radioactive waste disposal facility. Mixed wastes eligible for treatment include corrosive liquids (pH 12.5) containing EPA-regulated toxic metals (As, Ba, Pb, Cd, Cr, Ag, Se, Hg) at concentrations greater than the RCRA Toxicity Characteristic Leaching Procedure (TCLP) limit. The system has also been used to treat corrosive wastes with small quantities of fissionable materials. The AMWTS is a significant engineered solution with many improvements over the more labor intensive on-site treatment method being performed within a ventilation hood used previously. The previously used treatment system allowed for batch sizes of only 15-20 gallons whereas the new AMWTS allows for the treatment of batches up to 75 gallons; thereby reducing batch labor and supply costs by 40-60% and reducing analytical

Chemical composition analysis and product consistency tests to support Enhanced Hanford Waste Glass Models. Results for the Augusta and October 2014 LAW Glasses

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Best, D. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-07-07

In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for several simulated low activity waste (LAW) glasses (designated as the August and October 2014 LAW glasses) fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions.

Waste treatment

International Nuclear Information System (INIS)

Hutson, G.V.

1996-01-01

Numerous types of waste are produced by the nuclear industry ranging from high-level radioactive and heat-generating, HLW, to very low-level, LLW and usually very bulky wastes. These may be in solid, liquid or gaseous phases and require different treatments. Waste management practices have evolved within commercial and environmental constraints resulting in considerable reduction in discharges. (UK)

The role and results of the European Community's R and D work on radioactive waste management

International Nuclear Information System (INIS)

Orlowski, S.; Girardi, F.

1985-01-01

The titles of R and D programmes generally relate to a scientific discipline, a technology or a project: biotechnology, nuclear fission, etc. This is not so in the case of radioactive wastes, where R and D is focused on the management aspect. The role of R and D in general, and the contribution made by the Community programme in particular, are described and discussed with this in mind. Community R and D in the field of radioactive waste emerges as a powerful tool for establishing a broad consensus on delicate scientific questions such as the feasibility and long-term safety of the final storage of high activity wastes. Such a consensus is based on the many results obtained jointly by Community research teams over the last ten years. The implementation of three projects concerning experimental underground facilities in the context of the Community's new five-year (1985-1989) programme will provide the additional information that is needed before the large industrial disposal facilities of the future can be built

Surry Power Station, Units 1 and 2. Semiannual operating report, July--December 1974

International Nuclear Information System (INIS)

1974-01-01

Net electric power generated by Surry Unit 1 was 6,930,353 MWH with the generator on line for 10,417.7 hours. Net electric power generated by Unit 2 was 5,699,299 MWH with the generator on line for 8,384.2 hours. Information is presented concerning operation, radioactive effluent releases, solid radioactive wastes, fuel shipments, occurrences in which temperature limitations on the condenser cooling water discharge were exceeded, changes in station organization, occupational personnel radiation exposure, nonradiological monitoring including thermal, physical, and biological programs, and the radiological environmental monitoring program. (U.S.)

Hanford Waste Vitrification Plant Dangerous Waste Permit Application

International Nuclear Information System (INIS)

1991-10-01

The Hanford Facility currently stores mixed waste, resulting from various processing operations, in underground storage tanks. The Hanford Waste Vitrification Plant will be constructed and operated to process the high-activity fraction of mixed waste stored in these underground tanks. The Hanford Waste Vitrification Plant will solidify pretreated tank waste into a glass product that will be packaged for disposal in a national repository. This Vitrification Plant Dangerous Waste Permit Application, Revision 2, consists of both a Part A and a Part B permit application. An explanation of the Part A revisions, including Revision 4 submitted with this application, is provided at the beginning of the Part A section. The Part B consists of 15 chapters addressing the organization and content of the Part B Checklist prepared by the Washington State Department of Ecology (Ecology 1987)

Radioactive waste storage in mined caverns in crystalline rock: results of field investigations at Stripa, Sweden

International Nuclear Information System (INIS)

Witherspoon, P.A.

1980-10-01

It is generally agreed that the most practicable method of isolating nuclear wastes from the biosphere is by deep burial in suitable geologic formations. Such burial achieves a high degree of physical isolation but raises questions concerning the rate at which some of these wastes may return to the biosphere through transport by groundwater. Any suitable repository site will be disturbed first by excavation and second by the thermal pulse caused by the radioactive decay of the wastes. To assess the effectiveness of geologic isolation it is necessary to develop the capability of predicting the response of a rock mass to such a thermal pulse. Ultimately, this requires field measurements below the surface in media representative of those likely to be encountered at an actual repository. Access to a granitic rock mass adjacent to a defunct iron ore mine at Stripa, Sweden, at a depth of about 350 m below surface has provided a unique opportunity to conduct a comprehensive suite of hydrological and thermo-mechanical experiments under such conditions. The results of these field tests have shown the importance of geologic structure and the functional dependence of the thermo-mechanical properties on temperature in developing a valid predictive model. The results have also demonstrated the vital importance of carrying out large-scale investigations in a field test facility

Business unusual - Waste Act implementation: solid waste

CSIR Research Space (South Africa)

Oelofse, Suzanna HH

2013-08-01

Full Text Available The preamble to the Waste Act (2008) is very clear that, as a result of this legislation, waste management in South Africa will never be the same again. This should send a clear message that ‘business as usual’ will no longer be sufficient....

Waste processing practices at waste management department from INR

International Nuclear Information System (INIS)

Bujoreanu, D.; Bujoreanu, L.

2010-01-01

The Institute for Nuclear Research Pitesti (INR), subsidiary of the Romanian Authority for Nuclear Activities has its own Radioactive Waste Treatment Plant (STDR). The object of activity of STDR within the INR Pitesti is to treat and condition radioactive waste resulted from the nuclear facility. Also, it will must prepare and manage the decommissioning projects of its own facilities and to upgrade the facilities for the management of the radioactive waste resulting from other decommissioning activities. In according with the National Nuclear Program and the Governmental order no. 11/2003, the Institute for Nuclear Research is the main support for implementation of the methods and technologies for conditioning and disposal of radioactive waste generated by the decommissioning of nuclear facilities. The classes and criteria of classification for radioactive waste generated in operation and decommissioning in Romania are established in compliance with the classification recommended by IAEA and generally valid in EU countries. The general classification takes into consideration the disposal requirements to isolate the radioactive waste from environment. In Romania, waste minimization is considered by Order No. 56/2004 of CNCAN President for approval of Fundamental regulations on the safe management of radioactive waste. According to this regulation, the generation of radioactive waste is to be kept to the minimum practicable level in terms of both its activity and volume through appropriate design measures, facility operation and decommissioning practices. In order to meet this requirement, the operator must ensure: - selection and control of materials; - recycling and reuse of materials, including clearance of materials; - implementing adequate operating procedures, including those referring to the physical, chemical and radiological characterization of the waste and sorting of different type of materials. (orig.)

Sealing a nuclear waste repository in Columbia river basalt: preliminary results

International Nuclear Information System (INIS)

Hodges, F.N.

1980-01-01

The long containment time required of repositories for nuclear waste (10 4 to 10 6 years) requires that materials used for repository seals be stable in the geologic environment of the repository and of proven longevity. A list of candidate materials for sealing a repository in Columbia River Basalts has been prepared and refined through laboratory testing. The most feasible techniques for emplacing preferred plug materials have been identified and the resultant plugs have been evaluated on the basis of design functions. Preconceptual designs for tunnel, shaft, and borehole seals consist of multiple zone plugs with each zone fulfilling one or more design functions. Zones of disturbed rock around tunnels and shafts, resulting from excavation and subsequent stress release, are zones of higher permeability and of possible fluid migration. In preliminary designs the disturbed zones are blocked by cut-off collars filled with low permeability materials

A convenient method for estimating the contaminated zone of a subsurface aquifer resulting from radioactive waste disposal into ground

International Nuclear Information System (INIS)

Fukui, Masami; Katsurayama, Kousuke; Uchida, Shigeo.

1981-01-01

Studies were conducted to estimate the contamination spread resulting from the radioactive waste disposal into a subsurface aquifer. A general equation, expressing the contaminated zone as a function of radioactive decay, the physical and chemical parameters of soil is presented. A distribution coefficient was also formulated which can be used to judge the suitability of a site for waste disposal. Moreover, a method for predicting contaminant concentration in groundwater at a site boundary is suggested for a heterogeneous media where the subsurface aquifer has different values of porosity, density, flow velocity, distribution coefficient and so on. A general equation was also developed to predict the distribution of radionuclides resulting from the disposal of a solid waste material. The distributions of contamination was evaluated for 90 Sr and 239 Pu which obey a linear adsorption model and a first order kinetics respectively. These equations appear to have practical utility for easily estimating groundwater contamination. (author)

Attenuation of heavy metal leaching from hazardous wastes by co-disposal of wastes

Energy Technology Data Exchange (ETDEWEB)

Bae, Wookeun; Shin, Eung Bai [Hanyang Univ., Ansan (Korea, Republic of); Lee, Kil Chul; Kim, Jae Hyung [National Institute of Environmental Research, Seoul (Korea, Republic of)] [and others

1996-12-31

The potential hazard of landfill wastes was previously evaluated by examining the extraction procedures for individual waste, although various wastes were co-disposed of in actual landfills. This paper investigates the reduction of extraction-procedure toxicity by co-disposing various combinations of two wastes. When two wastes are mixed homogeneously, the extraction of heavy metals from the waste mixture is critically affected by the extract pH. Thus, co-disposal wastes will have a resultant pH between the pH values of its constituent. The lower the resultant pH, the lower the concentrations of heavy metals in the extract. When these wastes are extracted sequentially, the latter extracted waste has a stronger influence on the final concentration of heavy metals in the extract. Small-scale lysimeter experiments confirm that when heavy-metal-bearing leachates Generated from hazardous-waste lysimeters are passed through a nonhazardous-waste lysimeter filled with compost, briquette ash, or refuse-incineration ashes, the heavy-metal concentration in the final leachates decreases significantly. Thus, the heavy-metal leaching could be attenuated if a less extraction-procedure-toxic waste were placed at the bottom of a landfill. 3 refs., 4 figs., 5 tabs.

Results of detailed ground geophysical surveys for locating and differentiating waste structures in waste management area 'A' at Chalk River Laboratories, Ontario

Energy Technology Data Exchange (ETDEWEB)

Tomsons, D.K.; Street, P.J.; Lodha, G.S

1999-07-01

Waste Management Area 'A' (WMA 'A'), located in the outer area of the Chalk River Laboratories (CRL) was in use as a waste burial site from 1946 to 1955. Waste management structures include debris-filled trenches, concrete bunkers and miscellaneous contaminated solid materials, and ditches and pits used for liquid dispersal. In order to update historical records, it was proposed to conduct detailed ground geophysical surveys to define the locations of waste management structures in WMA 'A', assist in planning of the drilling and sampling program to provide ground truth for the geophysics investigation and to predict the nature and locations of unknown/undefined shallow structures. A detailed ground geophysical survey grid was established with a total of 127 grid lines, oriented NNE and spaced one metre apart. The geophysical surveys were carried out during August and September, 1996. The combination of geophysical tools used included the Geonics EM61 metal detector, the GSM-19 magnetometer/gradiometer and a RAMAC high frequency ground penetrating radar system. The geophysical surveys were successful in identifying waste management structures and in characterizing to some extent, the composition of the waste. The geophysical surveys are able to determine the presence of most of the known waste management structures, especially in the western and central portions of the grid which contain the majority of the metallic waste. The eastern portion of the grid has a completely different geophysical character. While historical records show that trenches were dug, they are far less evident in the geophysical record. There is clear evidence for a trench running between lines 30E and 63E at 70 m. There are indications from the radar survey of other trench-like structures in the eastern portion. EM61 data clearly show that there is far less metallic debris in the eastern portion. The geophysical surveys were also successful in identifying

Development of radiological performance objectives interim results: trade-offs in attitudes toward radioactive waste

International Nuclear Information System (INIS)

Lathrop, J.W.

1978-07-01

In order to measure the risk associated with radioactive waste it is necessary to ascertain public opinion concerning the relative significance of the different possible health effects of radiation, and public attitudes towards uncertainty. LLL has directed Decisions and Designs, Incorporated (DDI), to elicit such views from various members of the public. Purpose of this note is to give a brief account of some of the views so far obtained, provide some interpretation of these results, and briefly demonstrate how these results can be used to guide the drafting of regulations

Process Testing Results and Scaling for the Hanford Waste Treatment and Immobilization Plant (WTP) Pretreatment Engineering Platform - 10173

International Nuclear Information System (INIS)

Kurath, Dean E.; Daniel, Richard C.; Baldwin, David L.; Rapko, Brian M.; Barnes, Steven M.; Gilbert, Robert A.; Mahoney, Lenna A.; Huckaby, James L.

2010-01-01

The U.S. Department of Energy-Office of River Protections Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being designed and built to pretreat and then vitrify a large portion of the wastes in Hanfords 177 underground waste storage tanks at Richland, Washington. In support of this effort, engineering-scale tests at the Pretreatment Engineering Platform (PEP) have been completed to confirm the process design and provide improved projections of system capacity. The PEP is a 1/4.5-scale facility designed, constructed, and operated to test the integrated leaching and ultrafiltration processes being deployed at the WTP. The PEP replicates the WTP leaching processes with prototypic equipment and control strategies and non-prototypic ancillary equipment to support the core processing. The testing approach used a nonradioactive aqueous slurry simulant to demonstrate the unit operations of caustic and oxidative leaching, cross-flow ultrafiltration solids concentration, and solids washing. Parallel tests conducted at the laboratory scale with identical simulants provided results that allow scale-up factors to be developed between the laboratory and PEP performance. This paper presents the scale-up factors determined between the laboratory and engineering-scale results and presents arguments that extend these results to the full-scale process.

Waste management

International Nuclear Information System (INIS)

Soule, H.F.

1975-01-01

Current planning for the management of radioactive wastes, with some emphasis on plutonium contaminated wastes, includes the provision of re-positories from which the waste can be safely removed to permanent disposal. A number of possibilities for permanent disposal are under investigation with the most favorable, at the present time, apparently disposal in a stable geological formation. However, final choice cannot be made until all studies are completed and a pilot phase demonstrates the adequacy of the chosen method. The radioactive wastes which result from all portions of the fuel cycle could comprise an important source of exposure to the public if permitted to do so. The objectives of the AEC waste management program are to provide methods of treating, handling and storing these wastes so that this exposure will not occur. This paper is intended to describe some of the problems and current progress of waste management programs, with emphasis on plutonium-contaminated wastes. Since the technology in this field is advancing at a rapid pace, the descriptions given can be regarded only as a snapshot at one point in time. (author)

Transportation ALARA analysis for a nuclear waste management system

International Nuclear Information System (INIS)

McNair, G. W.; Schneider, K.; Smith, R.I.; Ross, W.; Faletti, D.

1988-01-01

In planning for implementation of a safe and cost-effective transportation system, the Department of Energy (DOE) commissioned the Pacific Northwest Laboratory (PNL) to develop estimates of the radiation doses, both public and occupational, that would result from operation of a system postulated using current designs and practices. From that evaluation, PNL identified activities/operations that result in relatively high doses, proposed conceptual alternatives that would effectively reduce such exposures, and evaluated the cost-effectiveness of such alternatives. This study contains an analysis of routine operations and estimates of the public and worker radiation doses that would occur in a postulated generic reference spent fuel transportation system using both truck and rail modes. Total risks are not estimated (i.e., consideration of nonradiological or accident risks that will be the subject of future studies in the transportation systems study plan 9TSSP) are not included). The system encompasses spent fuel loading at the reactor, transportation of the fuel to and from a receiving and handling facility and unloading of the fuel at a repository. The analysis provides cost/dose trade-offs of the postulated reference system as well as selected potential alternatives to the transportation system

Mixed Waste Focus Area - Waste form initiative

International Nuclear Information System (INIS)

Nakaoka, R.; Waters, R.; Pohl, P.; Roach, J.

1998-01-01

The mission of the US Department of Energy's (DOE) Mixed Waste Focus Area (MWFA) is to provide acceptable technologies that enable implementation of mixed waste treatment systems which are developed in partnership with end-users, stakeholders, tribal governments, and regulators. To accomplish this mission, a technical baseline was established in 1996 and revised in 1997. The technical baseline forms the basis for determining which technology development activities will be supported by the MWFA. The primary attribute of the technical baseline is a set of prioritized technical deficiencies or roadblocks related to implementation of mixed waste treatment systems. The Waste Form Initiative (WFI) was established to address an identified technical deficiency related to waste form performance. The primary goal of the WFI was to ensure that the mixed low-level waste (MLLW) treatment technologies being developed, currently used, or planned for use by DOE would produce final waste forms that meet the waste acceptance criteria (WAC) of the existing and/or planned MLLW disposal facilities. The WFI was limited to an evaluation of the disposal requirements for the radioactive component of MLLW. Disposal requirements for the hazardous component are dictated by the Resource Conservation and Recovery Act (RCRA), and were not addressed. This paper summarizes the technical basis, strategy, and results of the activities performed as part of the WFI

Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification

Energy Technology Data Exchange (ETDEWEB)

Saslow, Sarah A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Asmussen, Robert M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sahajpal, Rahul [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2017-07-01

This report describes the results from liquid secondary waste grout (LSWG) formulation and cementitious waste form qualification tests performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). New formulations for preparing a cementitious waste form from a high-sulfate liquid secondary waste stream simulant, developed for Effluent Management Facility (EMF) process condensates merged with low activity waste (LAW) caustic scrubber, and the release of key constituents (e.g. ⁹⁹Tc and ¹²⁹I) from these monoliths were evaluated. This work supports a technology development program to address the technology needs for Hanford Site Effluent Treatment Facility (ETF) liquid secondary waste (LSW) solidification and supports future Direct Feed Low-Activity Waste (DFLAW) operations. High-priority activities included simulant development, LSWG formulation, and waste form qualification. The work contained within this report relates to waste form development and testing and does not directly support the 2017 integrated disposal facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY17, and for future waste form development efforts. The provided data should be used by (i) cementitious waste form scientists to further understanding of cementitious dissolution behavior, (ii) IDF PA modelers who use quantified constituent leachability, effective diffusivity, and partitioning coefficients to advance PA modeling efforts, and (iii) the U.S. Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program. The results obtained help fill existing data gaps, support final selection of a LSWG waste form, and improve the technical defensibility of long-term waste form performance estimates.

Waste management - sewage - special wastes

International Nuclear Information System (INIS)

1987-01-01

The 27 papers represent a cross-section of the subject waste management. Particular attention is paid to the following themes: waste avoidance, waste product utilization, household wastes, dumping technology, sewage sludge treatments, special wastes, seepage from hazardous waste dumps, radioactive wastes, hospital wastes, purification of flue gas from waste combustion plants, flue gas purification and heavy metals, as well as combined sewage sludge and waste product utilization. The examples given relate to plants in Germany and other European countries. 12 papers have been separately recorded in the data base. (DG) [de

Chemical composition analysis and product consistency tests to support enhanced Hanford waste glass models: Results for the January, March, and April 2015 LAW glasses

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Riley, W. T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Best, D. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-03

In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for several simulated low activity waste (LAW) glasses (designated as the January, March, and April 2015 LAW glasses) fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions.

Mechanical compaction of Waste Isolation Pilot Plant simulated waste

International Nuclear Information System (INIS)

Butcher, B.M.; Thompson, T.W.; VanBuskirk, R.G.; Patti, N.C.

1991-06-01

The investigation described in this report acquired experimental information about how materials simulating transuranic (TRU) waste compact under axial compressive stress, and used these data to define a model for use in the Waste Isolation Pilot Plant (WIPP) disposal room analyses. The first step was to determine compaction curves for various simultant materials characteristic of TRU waste. Stress-volume compaction curves for various combinations of these materials were than derived to represent the combustible, metallic, and sludge waste categories. Prediction of compaction response in this manner is considered essential for the WIPP program because of the difficulties inherent in working with real (radioactive) waste. Next, full-sized 55-gallon drums of simulated combustible, metallic, and sludge waste were axially compacted. These results provided data that can be directly applied to room consolidation and data for comparison with the predictions obtained in Part 1 of the investigation. Compaction curves, which represent the combustible, metallic, and sludge waste categories, were determined, and a curve for the averaged waste inventory of the entire repository was derived. 9 refs., 31 figs., 12 tabs

Impacts of hazardous waste regulation on low-level waste management

International Nuclear Information System (INIS)

Sharples, F.E.; Eyman, L.D.

1987-01-01

Since passage of the 1984 amendments to the Resource Conservation and Recovery Act (RCRA), major changes have occurred in the regulation of hazardous waste. The US Environmental Protection Agency (EPA) has also greatly modified its interpretation of how these regulations apply to wastes from federal facilities, including defense wastes from US Department of Energy (DOE) sites. As a result, the regulatory distinctions between low-level radioactive waste (LLW) and hazardous waste are becoming blurred. This paper discusses recent statutory and regulatory changes and how they might affect the management of LLW at DOE facilities. 6 references

Nevada test site annual site environmental report for calendar year 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

Monitoring and surveillance on and around the Nevada Test Site (NTS) by US Department of Energy (DOE) contractors and NTS user organizations during 1995 indicated that operations on the NTS were conducted in compliance with applicable federal and DOE regulations and guidelines. All discharges of radioactive liquids remained onsite in containment ponds, and there was no indication of potential migration of radioactivity to the offsite area through groundwater. Surveillance around the NTS indicated that airborne radioactivity from diffusion, evaporation of effluents, or resuspension was not detectable offsite, and no measurable net exposure to members of the offsite population was detected through the offsite dosimetry program. There were no nonradiological releases to the offsite area. Hazardous wastes were shipped offsite to approved disposal facilities. Compliance with the various regulations stemming from the National Environmental Policy Act (NEPA) is being achieved and, where mandated, permits for air and water effluents and waste management have been obtained from the appropriate agencies. Cooperation with other agencies has resulted in seven different consent orders and agreements. Support facilities at off-NTS locations complied with the requirements of air quality permits and state or local wastewater discharge and hazardous waste permits.

Vitrification of Hanford wastes in a joule-heated ceramic melter and evaluation of resultant canisterized product

International Nuclear Information System (INIS)

Chapman, C.C.; Buelt, J.L.; Slate, S.C.; Katayama, Y.B.; Bunnell, L.R.

1979-08-01

Experience gained in the week-long vitrification test and characterization of the glass produced in the run support the following conclusions: The Hanford waste simulated in this test can be readily vitrified in a joule-heated ceramic melter. Physical properties of the molten glass were entirely compatible with melter operation. The average feed rate of 106 kg/h is high enough to make the ceramic melter a feasible piece of equipment for vitrifying Hanford wastes. The glass produced in this trial had good chemical durability, 6(10) -5 g/cm 2 -d. When one of the canisters was purposely dropped onto a steel pad, the damage was limited to deformation of the steel can in the impact area, cracking of a weld, and fracturing of glass in the immediate vicinity of the impact area. No glass was released from the canister as a result of the drop test. The results of this vitrification test support the technical feasibility of vitrifying Hanford wastes by means of a joule-heated ceramic melter. Surface area for large glass castings is equivalent to the mass median particle diameters between 4.27 cm (1.75 in.) and 8.91 cm (3.51 in.) even when allowed to cool rapidly by standing in ambient air. Large canisters (up to 0.91 m in dia) can be cast without large voids while standing in air if the fill rate is over 100 kg/h. 34 figures, 10 tables

The Radioactive Waste Management Advisory Committee's advice to ministers on the establishment of scientific consensus on the interpretation and significance of the results of science programmes into radioactive waste disposal

International Nuclear Information System (INIS)

1999-04-01

This document presents conclusions and recommendations on establishment of scientific consensus on the interpretation and significance of the results of science programmes into radioactive waste disposal. The topics discussed include: the nature of science and its limitations; societal views of science and the radioactive waste problem; issues upon which consensus will be needed; evidence of past attempts at greater involvement of the public; the linking of scientific and social consensus; communicating the nature of consensus to the public

Radioactive mixed waste disposal

International Nuclear Information System (INIS)

Jasen, W.G.; Erpenbeck, E.G.

1993-02-01

Various types of waste have been generated during the 50-year history of the Hanford Site. Regulatory changes in the last 20 years have provided the emphasis for better management of these wastes. Interpretations of the Atomic Energy Act of 1954 (AEA), the Resource Conservation and Recovery Act of 1976 (RCRA), and the Hazardous and Solid Waste Amendments (HSWA) have led to the definition of radioactive mixed wastes (RMW). The radioactive and hazardous properties of these wastes have resulted in the initiation of special projects for the management of these wastes. Other solid wastes at the Hanford Site include low-level wastes, transuranic (TRU), and nonradioactive hazardous wastes. This paper describes a system for the treatment, storage, and disposal (TSD) of solid radioactive waste

Thermally induced motion of marine sediments resulting from disposal of radioactive wastes

International Nuclear Information System (INIS)

Chavez, P.F.; Dawson, P.R.

1981-01-01

Coupled creep and heat transfer calculations have been performed to assess the sensitivity of heat load, viscosity, and canister density on the motion of waste canisters buried in marine sediments. Results indicate that no upward movement is predicted for heat loads remaining within the metallurgical and geochemical constraints placed on the temperature of sediments near the canister for the times analyzed. Upward movement of the canister is again not observed in calculations involving reasonable variations of the sediment viscosity and canister density. Maximum effective deviatoric stress levels due to thermally induced differential body forces are significantly less than the sediment's short term peak strength

Radioactive waste management

International Nuclear Information System (INIS)

Kawakami, Yutaka

2008-01-01

Radioactive waste generated from utilization of radioisotopes and each step of the nuclear fuel cycle and decommissioning of nuclear facilities are presented. On the safe management of radioactive waste management, international safety standards are established such as ''The Principles of Radioactive Waste Management (IAEA)'' and T he Joint Convention on the Safety of Radioactive Waste Management . Basic steps of radioactive waste management consist of treatment, conditioning and disposal. Disposal is the final step of radioactive waste management and its safety is confirmed by safety assessment in the licensing process. Safety assessment means evaluation of radiation dose rate caused by radioactive materials contained in disposed radioactive waste. The results of the safety assessment are compared with dose limits. The key issues of radioactive waste disposal are establishment of long term national strategies and regulations for safe management of radioactive waste, siting of repository, continuity of management activities and financial bases for long term, and security of human resources. (Author)

Vapor space characterization of waste tank 241-C-106: Results from samples collected on February 15, 1994

International Nuclear Information System (INIS)

McVeety, B.D.; Clauss, T.W.; Young, J.S.; Ligotke, M.W.; Goheen, S.C.; Lucke, R.B.; Pool, K.H.; McCulloch, M.; Fruchter, J.S.

1995-06-01

This document presents the details of the inorganic and organic analysis that was performed on samples from the headspace of Hanford waste tank 241-C-106. The results described were obtained to support the safety and toxicological evaluations. A summary of the results for the inorganic and organic analytes is included, as well as, a detailed description of the results which appears in the text

Management of radioactive waste

International Nuclear Information System (INIS)

Neerdael, B.; Marivoet, J.; Put, M.; Van Iseghem, P.; Volckaert, G.; Wacquier, W.

1998-09-01

The document gives an overview of of different aspects of radioactive waste management in Belgium. The document discusses the radioactive waste inventory in Belgium, the treatment and conditioning of radioactive waste as well as activities related to the characterisation of different waste forms. A separate chapter is dedicated to research and development regarding deep geological disposal of radioactive waste. In the Belgian waste management programme, particular emphasis is on studies for disposal in clay. Main results of these studies are highlighted and discussed

2005 dossier. ANDRA's researches on the geological disposal of high-level and long-lived radioactive wastes. Results and perspectives

International Nuclear Information System (INIS)

2005-06-01

This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the geologic disposal of high-level and long-lived radioactive wastes in deep geologic formations (argilites and granites). Content: 1 - Research on deep disposal of radioactive waste: general interest task: Legislative framework, ANDRA scientific objectives, Inspections and assessments; 2 - Designing a safe and reversible disposal system: Repository safety, Reversibility: an essential requirement; 3 - Clay Research on a repository in a clay formation, A long research programme, Dossier 2005 Argile; 4 - Meuse/Haute-Marne site clay: Expected properties of the rock formation, Choice of argillite, Meuse/Haute-Marne site, Conclusions from 10 years of research at the Meuse/Haute-Marne site; 5 - Repository installations: Safe and reversible architecture, Disposal of B waste, Disposal of C waste, Possible disposal of spent fuel (CU); 6 - The disposal facility in operation: From waste packages reception to their disposal in cells, Stages of the progressive closure of engineered structures; 7 - Reversible management: Freedom of choice for future generations, Various closure stages; 8 - Long-term evolution of the repository: Apprehending the repository complexity Main evolutions expected, Slow and limited release of radioactive substances; 9 - Repository safety and impact on man: Several evolution scenarios, Normal evolution, Altered evolution; 10 - Granite Research on a repository in a granite formation: A global approach, Scientific co-operations, Dossier 2005 Granite; 11 - Characteristics of French granite formations: What properties are required for a repository?, Different types of granite formations; 12 - Repository installations: Repository design adapted to granite fractures, Clay seals to prevent water flows, Waste disposal packages ensuring long-term leak-tightness, Physical and chemical environment favourable for waste packages, Architecture

Introduction of microbial nutrients in a nuclear fuel waste disposal vault as a result of excavation and operation activities

Energy Technology Data Exchange (ETDEWEB)

Stroes-Gascoyne, S; Gascoyne, M; Onagi, D; Thomas, D A; Hamon, C J; Watson, R; Porth, R J

1996-08-01

A nuclear fuel waste disposal vault would not likely be a sterile environment. Bacterial activity would be expected in those areas of the vault conducive to bacterial life, i.e., where effects of heat, moisture content, radiation and compaction would not prevent or severely restrict bacterial life and where suitable and sufficient nutrients would be present. An inventory of bacterial nutrients that would be emplaced `intentionally` with vault materials (fuel waste, waste containers, buffer and backfill materials) has been made previously. This report assesses bacterial nutrients that would be added `inadvertently` to a vault in the form of residues of materials used to excavate and operate a vault. Measurements of blasting material residues in the various water supplies, excavated broken rock (muck) and in cores drilled in old and new tunnel walls were made at AECL`s Underground Research Laboratory. Results show that the largest potential nutrient addition (both carbon and nitrogen) to a vault would result from using untreated excavated broken rock as part of the backfill. (author). 16 refs., 4 tabs., 10 figs.

Introduction of microbial nutrients in a nuclear fuel waste disposal vault as a result of excavation and operation activities

International Nuclear Information System (INIS)

Stroes-Gascoyne, S.; Gascoyne, M.; Onagi, D.; Thomas, D.A.; Hamon, C.J.; Watson, R.; Porth, R.J.

1996-08-01

A nuclear fuel waste disposal vault would not likely be a sterile environment. Bacterial activity would be expected in those areas of the vault conducive to bacterial life, i.e., where effects of heat, moisture content, radiation and compaction would not prevent or severely restrict bacterial life and where suitable and sufficient nutrients would be present. An inventory of bacterial nutrients that would be emplaced 'intentionally' with vault materials (fuel waste, waste containers, buffer and backfill materials) has been made previously. This report assesses bacterial nutrients that would be added 'inadvertently' to a vault in the form of residues of materials used to excavate and operate a vault. Measurements of blasting material residues in the various water supplies, excavated broken rock (muck) and in cores drilled in old and new tunnel walls were made at AECL's Underground Research Laboratory. Results show that the largest potential nutrient addition (both carbon and nitrogen) to a vault would result from using untreated excavated broken rock as part of the backfill. (author). 16 refs., 4 tabs., 10 figs

Final waste classification and waste form technical position papers

International Nuclear Information System (INIS)

1983-05-01

The waste classification technical position paper describes overall procedures acceptable to NRC staff which may be used by licensees to determine the presence and concentrations of the radionuclides listed in section 61.55, and thereby classifying waste for near-surface disposal. This technical position paper also provides guidance on the types of information which should be included in shipment manifests accompanying waste shipments to near-surface disposal facilities. The technical position paper on waste form provides guidance to waste generators on test methods and results acceptable to NRC staff for implementing the 10 CFR Part 61 waste form requirements. It can be used as an acceptable approach for demonstrating compliance with the 10 CFR Part 61 waste structural stability criteria. This technical position paper includes guidance on processing waste into an acceptable stable form, designing acceptable high-integrity containers, packaging cartridge filters, and minimizing radiation effects on organic ion-exchange resins. The guidance in the waste form technical position paper may be used by licensees as the basis for qualifying process control programs to meet the waste form stability requirements, including tests which can be used to demonstrate resistance to degradation arising from the effects of compression, moisture, microbial activity, radiation, and chemical changes. Generic test data (e.g., topical reports prepared by vendors who market solidification technology) may be used for process control program qualification where such generic data is applicable to the particular types of waste generated by a licensee

The UK waste input-output table: Linking waste generation to the UK economy.

Science.gov (United States)

Salemdeeb, Ramy; Al-Tabbaa, Abir; Reynolds, Christian

2016-10-01

In order to achieve a circular economy, there must be a greater understanding of the links between economic activity and waste generation. This study introduces the first version of the UK waste input-output table that could be used to quantify both direct and indirect waste arisings across the supply chain. The proposed waste input-output table features 21 industrial sectors and 34 waste types and is for the 2010 time-period. Using the waste input-output table, the study results quantitatively confirm that sectors with a long supply chain (i.e. manufacturing and services sectors) have higher indirect waste generation rates compared with industrial primary sectors (e.g. mining and quarrying) and sectors with a shorter supply chain (e.g. construction). Results also reveal that the construction, mining and quarrying sectors have the highest waste generation rates, 742 and 694 tonne per £1m of final demand, respectively. Owing to the aggregated format of the first version of the waste input-output, the model does not address the relationship between waste generation and recycling activities. Therefore, an updated version of the waste input-output table is expected be developed considering this issue. Consequently, the expanded model would lead to a better understanding of waste and resource flows in the supply chain. © The Author(s) 2016.

Food Waste Generation at Household Level: Results of a Survey among Employees of Two European Research Centers in Italy and Germany

Directory of Open Access Journals (Sweden)

Juliane Jörissen

2015-03-01

Full Text Available There is a broad consensus in literature that private households are significant contributors to the total amount of food waste in the EU. Thus, any strategy to meaningfully combat food wastage must put the end consumer in the center of prevention activities. This requires deeper insights into people’s motivations to discard still edible food and knowledge about potential barriers to reduce wasting. This paper reports on results of an online survey among two European research centers in Italy (JRC/Ispra and Germany (KIT/Karlsruhe. The focus of the survey was on households’ behaviors (shopping, eating, and food preparation habits and its influence on the generation of food waste. Furthermore, reasons for the disposal of food as well as measures and technologies most needed to prevent wastage were discussed. The results of the survey are analyzed, especially with regard to two questions: (1 Are there considerable differences between Ispra and Karlsruhe? (2 Are there considerable similarities or inconsistencies with the results of previous studies?

Information base for waste repository design. Volume 3. Waste/rock interactions

International Nuclear Information System (INIS)

Koplick, C.M.; Pentz, D.L.; Oston, S.G.; Talbot, R.

1979-01-01

This report describes the important effects resulting from interaction between radioactive waste and the rock in a nuclear waste repository. The state of the art in predicting waste/rock interactions is summarized. Where possible, independent numerical calculations have been performed. Recommendations are made pointing out areas which require additional research

Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

International Nuclear Information System (INIS)

Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S.K.; Vienna, John D.

2010-01-01

In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S. K.; Vienna, John D.

2010-08-01

In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

Ferrocyanide tank waste stability

International Nuclear Information System (INIS)

Fowler, K.D.

1993-01-01

Ferrocyanide wastes were generated at the Hanford Site during the mid to late 1950s as a result of efforts to create more tank space for the storage of high-level nuclear waste. The ferrocyanide process was developed to remove 137 CS from existing waste and newly generated waste that resulted from the recovery of valuable uranium in Hanford Site waste tanks. During the course of research associated with the ferrocyanide process, it was recognized that ferrocyanide materials, when mixed with sodium nitrate and/or sodium nitrite, were capable of violent exothermic reaction. This chemical reactivity became an issue in the 1980s, when safety issues associated with the storage of ferrocyanide wastes in Hanford Site tanks became prominent. These safety issues heightened in the late 1980s and led to the current scrutiny of the safety issues associated with these wastes, as well as current research and waste management programs. Testing to provide information on the nature of possible tank reactions is ongoing. This document supplements the information presented in Summary of Single-Shell Tank Waste Stability, WHC-EP-0347, March 1991 (Borsheim and Kirch 1991), which evaluated several issues. This supplement only considers information particular to ferrocyanide wastes

The structure of the Dutch waste sector and impediments for waste reduction

OpenAIRE

de Jong, P.; Wolsink, M.

1997-01-01

The way in which organizations collect, treat and dispose of waste in The Netherlands frustrates the achievement of waste reduction goals. The possibility that directed modification of the structure of the waste sector may contribute to stimulating consumers (i.e. all waste producers using services from collectors) to limit the generation of waste at the source by means of source reduction, re-use and recycling, is the subject of research of which the first results are presented here. This ar...

High Efficiency Particulate Air (HEPA) Filter Generation, Characterization, and Disposal Experiences at the Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Coffey, D. E.

2002-01-01

High Efficiency Particulate Air filtration is an essential component of the containment and ventilation systems supporting the research and development activities at the Oak Ridge National Laboratory. High Efficiency Particulate Air filters range in size from 7.6cm (3 inch) by 10.2 cm (4 inch) cylindrical shape filters to filter array assemblies up to 2.1 m (7 feet) high by 1.5 m (5 feet) wide. Spent filters are grouped by contaminates trapped in the filter media and become one of the components in the respective waste stream. Waste minimization and pollution prevention efforts are applied for both radiological and non-radiological applications. Radiological applications include laboratory hoods, glove boxes, and hot cells. High Efficiency Particulate Air filters also are generated from intake or pre-filtering applications, decontamination activities, and asbestos abatement applications. The disposal avenues include sanitary/industrial waste, Resource Conservation and Recovery Act and Toxic Substance Control Act, regulated waste, solid low-level waste, contact handled transuranic, and remote handled transuranic waste. This paper discusses characterization and operational experiences associated with the disposal of the spent filters across multiple applications

A comparative assessment of alternative waste management procedures for selected reprocessing wastes

International Nuclear Information System (INIS)

Hickford, G.E.; Plews, M.J.

1983-07-01

This report, which has been prepared by Associated Nuclear Services for the Department of the Environment, presents the results of a study and comparative assessment of management procedures for low and intermediate level solid waste streams arising from current and future fuel reprocessing operations on the Sellafield site. The characteristics and origins of the wastes under study are discussed and a reference waste inventory is presented, based on published information. Waste management strategy in the UK and its implications for waste conditioning, packaging and disposal are discussed. Wastes currently arising which are not suitable for Drigg burial or sea dumping are stored in an untreated form. Work is in hand to provide additional and improved disposal facilities which will accommodate all the waste streams under study. For each waste stream viable procedures are identified for further assessment. The procedures comprise a series of on-site operations-recovery from storage, pre-treatment, treatment, encapsulation, and packaging, prior to storage or disposal of the conditioned waste form. Assessments and comparisons of each procedure for each waste are presented. These address various process, operational, economic, radiological and general safety factors. The results are presented in a series of tables with supporting text. For the majority of wastes direct encapsulation with minimal treatment appears to be a viable procedure. Occupational exposure and general safety are not identified as significant factors governing the choice of procedures. The conditioned wastes meet the general requirements for safe handling during storage and transportation. The less active wastes suitable for disposal by currently available routes meet the appropriate disposal criteria. It is not possible to consider in detail the suitability for disposal of the more active wastes for which disposal facilities are not yet available. (Author)

Tank waste treatment science

International Nuclear Information System (INIS)

LaFemina, J.P.; Blanchard, D.L.; Bunker, B.C.; Colton, N.G.; Felmy, A.R.; Franz, J.A.; Liu, J.; Virden, J.W.

1994-01-01

Remediation efforts at the U.S. Department of Energy's Hanford Site require that many technical and scientific principles be combined for effectively managing and disposing the variety of wastes currently stored in underground tanks. Based on these principles, pretreatment technologies are being studied and developed to separate waste components and enable the most suitable treatment methods to be selected for final disposal of these wastes. The Tank Waste Treatment Science Task at Pacific Northwest Laboratory is addressing pretreatment technology development by investigating several aspects related to understanding and processing the tank contents. The experimental work includes evaluating the chemical and physical properties of the alkaline wastes, modeling sludge dissolution, and evaluating and designing ion exchange materials. This paper gives some examples of results of this work and shows how these results fit into the overall Hanford waste remediation activities. This work is part of series of projects being conducted for the Tank Waste Remediation System

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.

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

Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

Energy Technology Data Exchange (ETDEWEB)

Freudenreich, W.E.; Gruppelaar, H

1998-12-01

This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case {sup 99}Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k{sub eff}=0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The {sup 99} Tc-burner has a large initial loading; a more effective design may be possible. 5 refs.

Accelerator driven systems. ADS benchmark calculations. Results of stage 2. Radiotoxic waste transmutation

International Nuclear Information System (INIS)

Freudenreich, W.E.; Gruppelaar, H.

1998-12-01

This report contains the results of calculations made at ECN-Petten of a benchmark to study the neutronic potential of a modular fast spectrum ADS (Accelerator-Driven System) for radiotoxic waste transmutation. The study is focused on the incineration of TRans-Uranium elements (TRU), Minor Actinides (MA) and Long-Lived Fission Products (LLFP), in this case 99 Tc. The benchmark exercise is made in the framework of an IAEA Co-ordinated Research Programme. A simplified description of an ADS, restricted to the reactor part, with TRU or MA fuel (k eff =0.96) has been analysed. All spectrum calculations have been performed with the Monte Carlo code MCNP-4A. The burnup calculations have been performed with the code FISPACT coupled to MCNP-4A by means of our OCTOPUS system. The cross sections are based upon JEF-2.2 for transport calculations and supplemented with EAF-4 data for inventory calculations. The determined quantities are: core dimensions, fuel inventories, system power, sensitivity on external source spectrum and waste transmutation rates. The main conclusions are: The MA-burner requires only a small accelerator current increase during burnup, in contrast to the TRU-burner. The 99 Tc-burner has a large initial loading; a more effective design may be possible. 5 refs

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.

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

Headspace vapor characterization of Hanford waste tank 241-U-109: Results from samples collected on 8/10/95

International Nuclear Information System (INIS)

Evans, J.C.; Thomas, B.L.; Pool, K.H.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

1996-05-01

This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-U-109 (Tank U-109) At the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. This tank is on the Hydrogen Waste List. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company (WHC) and provided for analysis to Pacific Northwest National Laboratory (PNNL). Analyses were performed by the Vapor Analytical Laboratory (VAL) at PNNL. Analyte concentrations were based on analytical results and, where appropriate, sample volumes provided by WHC. A summary of the inorganic analytes, permanent gases and total non-methane hydrocarbons is listed in a table. The three highest concentration analytes detected in SUMMA trademark canister and triple sorbent trap samples is also listed in the table. Detailed descriptions of the analytical results appear in the text

Results of Toxicity Studies Conducted on Outfall X-08 and Its Contributing Waste Streams, November 1999 - June 2000

International Nuclear Information System (INIS)

Specht, W.L.

2000-01-01

This interim report summarizes the results of toxicity tests, Toxicity Identification Evaluations, and chemical analyses that have been conducted on SRS's NPDES Outfall X-08 and its contributing waste streams between November 1999 and June 2000

Waste management system alternatives for treatment of wastes from spent fuel reprocessing

International Nuclear Information System (INIS)

McKee, R.W.; Swanson, J.L.; Daling, P.M.

1986-09-01

This study was performed to help identify a preferred TRU waste treatment alternative for reprocessing wastes with respect to waste form performance in a geologic repository, near-term waste management system risks, and minimum waste management system costs. The results were intended for use in developing TRU waste acceptance requirements that may be needed to meet regulatory requirements for disposal of TRU wastes in a geologic repository. The waste management system components included in this analysis are waste treatment and packaging, transportation, and disposal. The major features of the TRU waste treatment alternatives examined here include: (1) packaging (as-produced) without treatment (PWOT); (2) compaction of hulls and other compactable wastes; (3) incineration of combustibles with cementation of the ash plus compaction of hulls and filters; (4) melting of hulls and failed equipment plus incineration of combustibles with vitrification of the ash along with the HLW; (5a) decontamination of hulls and failed equipment to produce LLW plus incineration and incorporation of ash and other inert wastes into HLW glass; and (5b) variation of this fifth treatment alternative in which the incineration ash is incorporated into a separate TRU waste glass. The six alternative processing system concepts provide progressively increasing levels of TRU waste consolidation and TRU waste form integrity. Vitrification of HLW and intermediate-level liquid wastes (ILLW) was assumed in all cases

Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

International Nuclear Information System (INIS)

Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

2010-01-01

In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.(1) The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

Energy Technology Data Exchange (ETDEWEB)

Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

2010-09-23

In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

BIOREFINE-2G — Result In Brief: Novel biopolymers from biorefinery waste-streams

DEFF Research Database (Denmark)

Stovicek, Vratislav; Chen, Xiao; Borodina, Irina

Second generation biorefineries are all about creating value from waste, so it seems only right that the ideal plant should leave nothing behind. With this in mind, the BIOREFINE-2G project has developed novel processes to convert pentose-rich side-streams into biopolymers.......Second generation biorefineries are all about creating value from waste, so it seems only right that the ideal plant should leave nothing behind. With this in mind, the BIOREFINE-2G project has developed novel processes to convert pentose-rich side-streams into biopolymers....

Management of radioactive wastes

International Nuclear Information System (INIS)

Hendee, W.R.

1984-01-01

The disposal of radioactive wastes is perhaps the most controversial and least understood aspect of the use of nuclear materials in generating electrical power, the investigation of biochemical processes through tracer kinetics, and the diagnosis and treatment of disease. In the siting of nuclear power facilities, the disposal of radioactive wastes is invariably posed as the ultimate unanswerable question. In the fall of 1979, biochemical and physiologic research employing radioactive tracers was threatened with a slowdown resulting from temporary closure of sites for disposal of low-level radioactive wastes (LLW). Radioactive pharmaceuticals used extensively for diagnosis and treatment of human disease have increased dramatically in price, partly as a result of the escalating cost of disposing of radioactive wastes created during production of the labeled pharmaceuticals. These problems have resulted in identification of the disposal of LLW as the most pressing issue in the entire scheme of management of hazardous wastes. How this issue as well as the separate issue of disposal of high-level radioactive wastes (HLW) are being addressed at both national and state levels is the subject of this chapter

Radioactive wastes

International Nuclear Information System (INIS)

Dupuis, M.C.

2007-01-01

Managing radioactive wastes used to be a peripheral activity for the French atomic energy commission (Cea). Over the past 40 years, it has become a full-fledged phase in the fuel cycle of producing electricity from the atom. In 2005, the national radioactive waste management agency (ANDRA) presented to the government a comprehensive overview of the results drawn from 15 years of research. This landmark report has received recognition beyond France's borders. By broadening this agency's powers, an act of 28 June 2006 acknowledges the progress made and the quality of the results. It also sets an objective for the coming years: work out solutions for managing all forms of radioactive wastes. The possibility of recovering wastes packages from the disposal site must be assured as it was asked by the government in 1998. The next step will be the official demand for the creation of a geological disposal site in 2016

Regional waste treatment with monolith disposal for low-level radioactive waste

International Nuclear Information System (INIS)

Forsberg, C.W.

1983-01-01

An alternative system is proposed for the disposal of low-level radioactive waste. This system, called REgional Treatment with MOnolith Disposal (RETMOD), is based on integrating three commercial technologies: automated package warehousing, whole-barrel rotary kiln incineration, and cement-based grouts for radioactive waste disposal. In the simplified flowsheet, all the sludges, liquids, resins, and combustible wastes are transported to regional facilities where they are incinerated. The ash is then mixed with special cement-based grouts, and the resulting mixture is poured into trenches to form large waste-cement monoliths. Wastes that do not require treatment, such as damaged and discarded equipment, are prepositioned in the trenches with the waste-cement mixture poured on top. The RETMOD system may provide higher safety margins by conversion of wastes into a solidified low-leach form, creation of low-surface area waste-cement monoliths, and centralization of waste processing into a few specialized facilities. Institutional problems would be simplified by placing total responsibility for safe disposal on the disposal site operator. Lower costs may be realized through reduced handling costs, the economics of scale, simplified operations, and less restrictive waste packaging requirements

Stabilization of compactible waste

International Nuclear Information System (INIS)

Franz, E.M.; Heiser, J.H. III; Colombo, P.

1990-09-01

This report summarizes the results of series of experiments performed to determine the feasibility of stabilizing compacted or compactible waste with polymers. The need for this work arose from problems encountered at disposal sites attributed to the instability of this waste in disposal. These studies are part of an experimental program conducted at Brookhaven National Laboratory (BNL) investigating methods for the improved solidification/stabilization of DOE low-level wastes. The approach taken in this study was to perform a series of survey type experiments using various polymerization systems to find the most economical and practical method for further in-depth studies. Compactible dry bulk waste was stabilized with two different monomer systems: styrene-trimethylolpropane trimethacrylate (TMPTMA) and polyester-styrene, in laboratory-scale experiments. Stabilization was accomplished by wetting or soaking compactible waste (before or after compaction) with monomers, which were subsequently polymerized. Three stabilization methods are described. One involves the in-situ treatment of compacted waste with monomers in which a vacuum technique is used to introduce the binder into the waste. The second method involves the alternate placement and compaction of waste and binder into a disposal container. In the third method, the waste is treated before compaction by wetting the waste with the binder using a spraying technique. A series of samples stabilized at various binder-to-waste ratios were evaluated through water immersion and compression testing. Full-scale studies were conducted by stabilizing two 55-gallon drums of real compacted waste. The results of this preliminary study indicate that the integrity of compacted waste forms can be readily improved to ensure their long-term durability in disposal environments. 9 refs., 10 figs., 2 tabs

Properties of radioactive wastes and waste containers

International Nuclear Information System (INIS)

Arora, H.S.; Dayal, R.

1984-01-01

Major tasks in this NRC-sponsored program include: (1) an evaluation of the acceptability of low-level solidified wastes with respect to minimizing radionuclide releases after burial; and (2) an assessment of the influence of pertinent environmental stresses on the performance of high-integrity radwaste container (HIC) materials. The waste form performance task involves studies on small-scale laboratory specimens to predict and extrapolate: (1) leachability for extended time periods; (2) leach behavior of full-size forms; (3) performance of waste forms under realistic leaching conditions; and (4) leachability of solidified reactor wastes. The results show that leach data derived from testing of small-scale specimens can be extrapolated to estimate leachability of a full-scale specimen and that radionuclide release data derived from testing of simulants can be employed to predict the release behavior of reactor wastes. Leaching under partially saturated conditions exhibits lower releases of radionuclides than those observed under the conventional IAEA-type or ANS 16.1 leach tests. The HIC assessment task includes the characterization of mechanical properties of Marlex CL-100, a candidate radwaste high density polyethylene material. Tensile strength and creep rupture tests have been carried out to determine the influence of specific waste constituents as well as gamma irradiation on material performance. Emphasis in ongoing tests is being placed on studying creep rupture while the specimens are in contact with a variety of chemicals including radiolytic by-products of irradiated resin wastes. 12 references 6 figures, 2 tables

Postclosure safety assessment of a used fuel repository in sedimentary rock

International Nuclear Information System (INIS)

Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E.

2014-01-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. This paper summarizes an illustrative case study of the current multi-barrier design and postclosure safety of a deep geological repository in a hypothetical sedimentary Michigan Basin setting. The purpose of this postclosure safety assessment is to determine potential effects of the repository on the health and safety of persons and the environment. Results are compared against acceptance criteria established for the protection of persons and the environment from potential radiological and non-radiological hazards. (author)

Postclosure safety assessment of a used fuel repository in sedimentary rock

Energy Technology Data Exchange (ETDEWEB)

Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E. [Nuclear Waste Management Organization, Toronto, ON (Canada)

2014-07-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. This paper summarizes an illustrative case study of the current multi-barrier design and postclosure safety of a deep geological repository in a hypothetical sedimentary Michigan Basin setting. The purpose of this postclosure safety assessment is to determine potential effects of the repository on the health and safety of persons and the environment. Results are compared against acceptance criteria established for the protection of persons and the environment from potential radiological and non-radiological hazards. (author)

Waste Receiving and Processing (WRAP) Facility Weight Scale Analysis Fairbanks Weight Scale Evaluation Results

International Nuclear Information System (INIS)

JOHNSON, M.D.

1999-01-01

Fairbanks Weight Scales are used at the Waste Receiving and Processing (WRAP) facility to determine the weight of waste drums as they are received, processed, and shipped. Due to recent problems, discovered during calibration, the WRAP Engineering Department has completed this document which outlines both the investigation of the infeed conveyor scale failure in September of 1999 and recommendations for calibration procedure modifications designed to correct deficiencies in the current procedures

Radioactive wastes and discharges

International Nuclear Information System (INIS)

1993-01-01

According to the Section 24 of the Finnish Radiation Decree (1512/91), the Finnish Centre for Radiation and Nuclear Safety shall specify the concentration and activity limits and principles for the determination whether a waste can be defined as a radioactive waste or not. The radiation safety requirements and limits for the disposal of radioactive waste are given in the guide. They must be observed when discharging radioactive waste into the atmosphere or sewer system, or when delivering solid low-activity waste to a landfill site without a separate waste disposal plan. The guide does not apply to the radioactive waste resulting from the utilization of nuclear energy of natural resources. (4 refs., 1 tab.)

Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 2. Testing and results

International Nuclear Information System (INIS)

Hartwell, J.K.; McIlwain, M.E.

2005-01-01

Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. These tests and their results are described; while the former paper in this issue presents the methodology, equipment and techniques. (author)

Colloid Genesis/Transport and Flow Pathway Alterations Resulting From Interactions of Reactive Waste Solutions and Hanford Vadose Zone Sediments

International Nuclear Information System (INIS)

Wan, Jiamin; Tokunaga, Tetsu K.

2001-01-01

Leakage of underground tanks containing high-level nuclear waste solutions has been identified at various DOE facilities. The Hanford Site is one the main facilities of concern, with about 2,300 to 3,400 m3 of leaked waste liquids. Radionuclides and other contaminants have been found in elevated concentrations in the vadose zone and groundwater underneath single shell tank farms. We do not currently know the mechanisms responsible for the unexpected deep migration of some contaminants through the vadose zone, and such understanding is urgently needed for planning remediation. Due to the extreme chemical conditions of the tank waste solutions (very high pH, aluminum concentration, and ionic strength), interactions between the highly reactive waste solutions and sediments underneath the tanks can result in dissolution of primary minerals of the sediments and precipitation of secondary phases including colloidal particles. Contaminants can sorb onto and/or co-precipitate with the secondary phases. Therefore transport of strongly associated contaminants on mobile colloids can be substantially greater than without colloids. The overall objective of this research is to improve our understanding on the effects of interactions between the tank waste solution and sediments on deep contaminant migration under Hanford Site conditions. This objective will be achieved through the following four tasks: (1) colloid generation and transport studies, (2) studies on sediment permeability and chemical composition alterations, (3) quantifying associations of contaminants with secondary colloids, and (4) studies on the combined effects of the aforementioned processes on deep contaminant migration

Radioactive wastes and discharges

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

The guide sets out the radiation safety requirements and limits for the treatment of radioactive waste. They shall be observed when discharging radioactive substances into the atmosphere or sewer system, or when delivering solid, low-activity waste to a landfill site without a separate waste treatment plan. The guide does not apply to the radioactive waste resulting from the utilisation of nuclear energy or natural resources.

Radioactive wastes and discharges

International Nuclear Information System (INIS)

2000-01-01

The guide sets out the radiation safety requirements and limits for the treatment of radioactive waste. They shall be observed when discharging radioactive substances into the atmosphere or sewer system, or when delivering solid, low-activity waste to a landfill site without a separate waste treatment plan. The guide does not apply to the radioactive waste resulting from the utilisation of nuclear energy or natural resources

Glassy slags as novel waste forms for remediating mixed wastes with high metal contents

International Nuclear Information System (INIS)

Feng, X.; Wronkiewicz, D.J.; Bates, J.K.; Brown, N.R.; Buck, E.C.; Gong, M.; Ebert, W.L.

1994-01-01

Argonne National Laboratory (ANL) is developing a glassy slag final waste form for the remediation of low-level radioactive and mixed wastes with high metal contents. This waste form is composed of various crystalline and metal oxide phases embedded in a silicate glass phase. This work indicates that glassy slag shows promise as final waste form because (1) it has similar or better chemical durability than high-level nuclear waste (HLW) glasses, (2) it can incorporate large amounts of metal wastes, (3) it can incorporate waste streams having low contents of flux components (boron and alkalis), (4) it has less stringent processing requirements (e.g., viscosity and electric conductivity) than glass waste forms, (5) its production can require little or no purchased additives, which can result in greater reduction in waste volume and overall treatment costs. By using glassy slag waste forms, minimum additive waste stabilization approach can be applied to a much wider range of waste streams than those amenable only to glass waste forms

Effect of COSMOS technologies in detoxifying municipal solid waste incineration fly ash, preliminary results

Science.gov (United States)

Piccinelli, Elsa; Lasagni, Marina; Collina, Elena; Bonaiti, Stefania; Bontempi, Elza

2017-05-01

This study investigates the effect of technologies for heavy metal stabilization on the concentration of PolyChlorinatedDibenzo-p-Dioxins (PCDD) and PolyChlorinatedDibenzoFurans (PCDF), abbreviated PCDD/F, in Municipal Solid Waste Incineration (MSWI) fly ash. We determined the variation of the Total Organic Carbon (TOC) and PCDD/F concentration between raw and stabilized material. The technologies, that already proved to be very promising for heavy metal entrapment, showed encouraging results also for PCDD/F detoxification. This result could be very impacting on the management of MSWI fly ash: at the best of our knowledge, there are no methods, in literature, that can provide good results in stabilization of heavy metals, and abatement of chlorinated organic pollutants contained in the same matrix.

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

Defining a metal-based waste form for IFR pyroprocessing wastes

International Nuclear Information System (INIS)

McDeavitt, S.M.; Park, J.Y.; Ackerman, J.P.

1994-01-01

Pyrochemical electrorefining to recover actinides from metal nuclear fuel is a key element of the Integral Fast Reactor (IFR) fuel cycle. The process separates the radioactive fission products from the long-lived actinides in a molten LiCl-KCl salt, and it generates a lower waste volume with significantly less long-term toxicity as compared to spent nuclear fuel. The process waste forms include a mineral-based waste form that will contain fission products removed from an electrolyte salt and a metal-based waste form that will contain metallic fission products and the fuel cladding and process materials. Two concepts for the metal-based waste form are being investigated: (1) encapsulating the metal constituents in a Cu-Al alloy and (2) alloying the metal constituents into a uniform stainless steel-based waste form. Results are given from our recent studies of these two concepts

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

Property-close source separation of hazardous waste and waste electrical and electronic equipment - A Swedish case study

International Nuclear Information System (INIS)

Bernstad, Anna; Cour Jansen, Jes la; Aspegren, Henrik

2011-01-01

Through an agreement with EEE producers, Swedish municipalities are responsible for collection of hazardous waste and waste electrical and electronic equipment (WEEE). In most Swedish municipalities, collection of these waste fractions is concentrated to waste recycling centres where households can source-separate and deposit hazardous waste and WEEE free of charge. However, the centres are often located on the outskirts of city centres and cars are needed in order to use the facilities in most cases. A full-scale experiment was performed in a residential area in southern Sweden to evaluate effects of a system for property-close source separation of hazardous waste and WEEE. After the system was introduced, results show a clear reduction in the amount of hazardous waste and WEEE disposed of incorrectly amongst residual waste or dry recyclables. The systems resulted in a source separation ratio of 70 wt% for hazardous waste and 76 wt% in the case of WEEE. Results show that households in the study area were willing to increase source separation of hazardous waste and WEEE when accessibility was improved and that this and similar collection systems can play an important role in building up increasingly sustainable solid waste management systems.

Assessment of management alternatives for LWR wastes. Volume 5. Assessment of the radiological impact to the public resulting from discharges of radioactive effluents

International Nuclear Information System (INIS)

Centner, B.

1993-01-01

This report deals with the assessment of the radiological impact to the public resulting from discharges of radioactive effluents (liquid and gaseous) in connection with the implementation of the Belgian scenario for the management of PWR waste. Both individual and collective doses have been estimated for a critical group of the population living around the nuclear power plants concerned. This study is part of an overall theoretical exercise aimed at evaluating a selection of management wastes for LWR waste based on economical and radiological criteria

Fatal and nonfatal risk associated with recycle of D&D-generated concrete

Energy Technology Data Exchange (ETDEWEB)

Boren, J.K.; Ayers, K.W.; Parker, F.L. [Vanderbilt Univ., Nashville, TN (United States)

1997-02-01

As decontamination and decommissioning activities proceed within the U.S. Department of Energy Complex, vast volumes of uncontaminated and contaminated concrete will be generated. The current practice of decontaminating and landfilling the concrete is an expensive and potentially wasteful practice. Research is being conducted at Vanderbilt University to assess the economic, social, legal, and political ramifications of alternate methods of dealing with waste concrete. An important aspect of this research work is the assessment of risk associated with the various alternatives. A deterministic risk assessment model has been developed which quantifies radiological as well as non-radiological risks associated with concrete disposal and recycle activities. The risk model accounts for fatal as well as non-fatal risks to both workers and the public. Preliminary results indicate that recycling of concrete presents potentially lower risks than the current practice. Radiological considerations are shown to be of minor importance in comparison to other sources of risk, with conventional transportation fatalities and injuries dominating. Onsite activities can also be a major contributor to non-fatal risk.

Waste Sites - Municipal Waste Operations

Data.gov (United States)

NSGIC Education | GIS Inventory — A Municipal Waste Operation is a DEP primary facility type related to the Waste Management Municipal Waste Program. The sub-facility types related to Municipal Waste...

Solid waste study

International Nuclear Information System (INIS)

Ortiz, Paul G.

1995-01-01

The purpose of this document is to study the solid waste issues brought about by a Type C Investigation; ''Disposal of Inappropriate Material in the Los Alamos County Landfill'' (May 28, 1993). The study was completed in August 1995 by Coleman Research Corporation, under subcontract number 405810005-Y for Los Alamos National Laboratory (LANL). The study confirmed the issues identified in the Type C investigation, and also ascertained further issues or problems. During the course of this study two incidents involving hazardous waste resulted in the inappropriate disposal of the waste. An accidental spill, on June 8, 1995, at one of Laboratory buildings was not handled correctly, and ended up in the LAC Landfill. Hazardous waste was disposed of in a solid waste container and sent to the Los Alamos County Landfill. An attempt to locate the hazardous waste at the LAC Landfill was not successful. The second incident involving hazardous waste was discovered by the FSS-8, during a random dumpster surveillance. An interim dumpster program managed by FSS-8 discovered hazardous waste and copper chips in the solid waste, on August 9, 1995. The hazardous waste and copper chips would have been transported to the LAC Landfill if the audit team had not brought the problem to the awareness of the facility waste management personnel

SOLID WASTE STUDY

Energy Technology Data Exchange (ETDEWEB)

PAUL G. ORTIZ - COLEMAN RESEARCH CORP/COMPA INDUSTRIES

1995-08-01

The purpose of this document is to study the solid waste issues brought about by a Type C Investigation; ``Disposal of Inappropriate Material in the Los Alamos County Landfill'' (May 28, 1993). The study was completed in August 1995 by Coleman Research Corporation, under subcontract number 405810005-Y for Los Alamos National Laboratory (LANL). The study confirmed the issues identified in the Type C investigation, and also ascertained further issues or problems. During the course of this study two incidents involving hazardous waste resulted in the inappropriate disposal of the waste. An accidental spill, on June 8, 1995, at one of Laboratory buildings was not handled correctly, and ended up in the LAC Landfill. Hazardous waste was disposed of in a solid waste container and sent to the Los Alamos County Landfill. An attempt to locate the hazardous waste at the LAC Landfill was not successful. The second incident involving hazardous waste was discovered by the FSS-8, during a random dumpster surveillance. An interim dumpster program managed by FSS-8 discovered hazardous waste and copper chips in the solid waste, on August 9, 1995. The hazardous waste and copper chips would have been transported to the LAC Landfill if the audit team had not brought the problem to the awareness of the facility waste management personnel.

Stabilization and disposal of Argonne-West low-level mixed wastes in ceramicrete waste forms

International Nuclear Information System (INIS)

Barber, D. B.; Singh, D.; Strain, R. V.; Tlustochowicz, M.; Wagh, A. S.

1998-01-01

The technology of room-temperature-setting phosphate ceramics or Ceramicretetrademark technology, developed at Argonne National Laboratory (ANL)-East is being used to treat and dispose of low-level mixed wastes through the Department of Energy complex. During the past year, Ceramicretetrademark technology was implemented for field application at ANL-West. Debris wastes were treated and stabilized: (a) Hg-contaminated low-level radioactive crushed light bulbs and (b) low-level radioactive Pb-lined gloves (part of the MWIR number s ign AW-W002 waste stream). In addition to hazardous metals, these wastes are contaminated with low-level fission products. Initially, bench-scale waste forms with simulated and actual waste streams were fabricated by acid-base reactions between mixtures of magnesium oxide powders and an acid phosphate solution, and the wastes. Size reduction of Pb-lined plastic glove waste was accomplished by cryofractionation. The Ceramicretetrademark process produces dense, hard ceramic waste forms. Toxicity Characteristic Leaching Procedure (TCLP) results showed excellent stabilization of both Hg and Pb in the waste forms. The principal advantage of this technology is that immobilization of contaminants is the result of both chemical stabilization and subsequent microencapsulation of the reaction products. Based on bench-scale studies, Ceramicretetrademark technology has been implemented in the fabrication of 5-gal waste forms at ANL-West. Approximately 35 kg of real waste has been treated. The TCLP is being conducted on the samples from the 5-gal waste forms. It is expected that because the waste forms pass the limits set by the EPAs Universal Treatment Standard, they will be sent to a radioactive-waste disposal facility

Waste-clearance strategy for DOE waste processed at commercial facilities

International Nuclear Information System (INIS)

Chen, S.Y.; Pfingston, M.; LePoire, D.

1996-01-01

In May 1991, a moratorium was issued on shipping potentially mixed waste from DOE facilities nationwide to commercial treatment, storage, and disposal facilities. A potential waste-clearance strategy was developed to address the DOE mixed-waste moratorium issues, which had resulted from a lack of exisitng volume contamination regulations. This strategy also has important potential applications for establishing site clearance limits that ensure worker and public risks remain well below regulatory limits

A waste characterization monitor for low-level radioactive waste management

International Nuclear Information System (INIS)

Davey, E.C.; Csullog, G.W.; Kupca, S.; Hippola, K.B.

1985-06-01

The exploitation of nuclear processes and technology for the benefit of Canadians results in the routine generation of approximately 12 000 m 3 of solid low-level radioactive waste annually. To protect the public and the environment, this waste must be isolated for the duration of its potential hazard. In Canada, current planning foresees the development and use of a range of storage and disposal facilities exhibiting differing containment capabilities. To demonstrate adequate isolation safety and to minimize overall costs, the radionuclide content of waste items must be quantified so that the radiological hazards of each waste item can be matched to the isolation capabilities of specific containment facilities. This paper describes a non-invasive, waste characterization monitor that is capable of quantifying the radionuclide content of low-level waste packages to the 9 Bq/g (250 pCi/g) level. The assay technique is based on passive gamma-ray spectroscopy where the concentration of gamma-ray emitting radionuclides in a waste item can be estimated from the analysis of the gamma-ray spectra of the item and calibrated standards

Waste pipe calculus extensions

International Nuclear Information System (INIS)

O'Connell, W.J.

1979-01-01

The waste pipe calculus provides a rapid method, using Laplace transforms, to calculate the transport of a pollutant such as nuclear waste, by a network of one-dimensional flow paths. The present note extends previous work as follows: (1) It provides an alternate approximation to the time-domain function (inverse Laplace transform) for the resulting transport. This algebraic approximation may be viewed as a simpler and more approximate model of the transport process. (2) It identifies two scalar quantities which may be used as summary consequence measures of the waste transport (or inversely, waste retention) system, and provides algebraic expressions for them. (3) It includes the effects of radioactive decay on the scalar quantity results, and further provides simplifying approximations for the cases of medium and long half-lives. This algebraic method can be used for quick approximate analyses of expected results, uncertainty and sensitivity, in evaluating selection and design choices for nuclear waste disposal systems

Treatability study of absorbent polymer waste form for mixed waste treatment

International Nuclear Information System (INIS)

Herrmann, S. D.; Lehto, M. A.; Stewart, N. A.; Croft, A. D.; Kern, P. W.

2000-01-01

A treatability study was performed to develop and characterize an absorbent polymer waste form for application to low level (LLW) and mixed low level (MLLW) aqueous wastes at Argonne National Laboratory-West (ANL-W). In this study absorbent polymers proved effective at immobilizing aqueous liquid wastes in order to meet Land Disposal Restrictions for subsurface waste disposal. Treatment of aqueous waste with absorbent polymers provides an alternative to liquid waste solidification via high-shear mixing with clays and cements. Significant advantages of absorbent polymer use over clays and cements include ease of operations and waste volume minimization. Absorbent polymers do not require high-shear mixing as do clays and cements. Granulated absorbent polymer is poured into aqueous solutions and forms a gel which passes the paint filter test as a non-liquid. Pouring versus mixing of a solidification agent not only eliminates the need for a mixing station, but also lessens exposure to personnel and the potential for spread of contamination from treatment of radioactive wastes. Waste minimization is achieved as significantly less mass addition and volume increase is required of and results from absorbent polymer use than that of clays and cements. Operational ease and waste minimization translate into overall cost savings for LLW and MLLW treatment

Savannah River Site environmental report for 1993 summary pamphlet

International Nuclear Information System (INIS)

Karapatakis, L.

1994-01-01

This pamphlet summarizes the impact of 1993 Savannah River Site operations on the environment and the off-site public. It includes an overview of site operations; the basis for radiological and nonradiological monitoring; 1993 radiological releases and the resulting dose to the off-site population; and results of the 1993 nonradiological program. The Savannah River Site Environmental Report for 1993 describes the findings of the environmental monitoring program for 1993. The report contains detailed information about site operations,the environmental monitoring and surveillance programs, monitoring and surveillance results, environmental compliance activities, and special programs. The report is distributed to government officials, members of the US Congress, universities, government facilities, environmental and civic groups, the news media, and interested individuals

Determinants of consumer food waste behaviour: Two routes to food waste.

Science.gov (United States)

Stancu, Violeta; Haugaard, Pernille; Lähteenmäki, Liisa

2016-01-01

Approximately one quarter of the food supplied for human consumption is wasted across the food supply chain. In the high income countries, the food waste generated at the household level represents about half of the total food waste, making this level one of the biggest contributors to food waste. Yet, there is still little evidence regarding the determinants of consumers' food waste behaviour. The present study examines the effect of psycho-social factors, food-related routines, household perceived capabilities and socio-demographic characteristics on self-reported food waste. Survey data gathered among 1062 Danish respondents measured consumers' intentions not to waste food, planning, shopping and reuse of leftovers routines, perceived capability to deal with household food-related activities, injunctive and moral norms, attitudes towards food waste, and perceived behavioural control. Results show that perceived behavioural control and routines related to shopping and reuse of leftovers are the main drivers of food waste, while planning routines contribute indirectly. In turn, the routines are related to consumers' perceived capabilities to deal with household related activities. With regard to intentional processes, injunctive norms and attitudes towards food waste have an impact while moral norms and perceived behavioural control make no significant contribution. Implications of the study for initiatives aimed at changing consumers' food waste behaviour are discussed. Copyright © 2015 Elsevier Ltd. All rights reserved.

Efficiency of energy recovery from municipal solid waste and the resultant effect on the greenhouse gas balance.

Science.gov (United States)

Gohlke, Oliver

2009-11-01

Global warming is a focus of political interest and life-cycle assessment of waste management systems reveals that energy recovery from municipal solid waste is a key issue. This paper demonstrates how the greenhouse gas effects of waste treatment processes can be described in a simplified manner by considering energy efficiency indicators. For evaluation to be consistent, it is necessary to use reasonable system boundaries and to take the generation of electricity and the use of heat into account. The new European R1 efficiency criterion will lead to the development and implementation of optimized processes/systems with increased energy efficiency which, in turn, will exert an influence on the greenhouse gas effects of waste management in Europe. Promising technologies are: the increase of steam parameters, reduction of in-plant energy consumption, and the combined use of heat and power. Plants in Brescia and Amsterdam are current examples of good performance with highly efficient electricity generation. Other examples of particularly high heat recovery rates are the energy-from-waste (EfW) plants in Malmö and Gothenburg. To achieve the full potential of greenhouse gas reduction in waste management, it is necessary to avoid landfilling combustible wastes, for example, by means of landfill taxes and by putting incentives in place for increasing the efficiency of EfW systems.

Understanding the role of waste prevention in local waste management: A literature review.

Science.gov (United States)

Zacho, Kristina O; Mosgaard, Mette A

2016-10-01

Local waste management has so far been characterised by end-of-pipe solutions, landfilling, incineration, and recycling. End-of-pipe solutions build on a different mind-set than life cycle-based approaches, and for this reason, local waste managers are reluctant to consider strategies for waste prevention. To accelerate the transition of waste and resource management towards a more integrated management, waste prevention needs to play a larger role in the local waste management. In this review article, we collect knowledge from the scientific community on waste prevention of relevance to local waste management. We analyse the trends in the waste prevention literature by organising the literature into four categories. The results indicate an increasing interest in waste prevention, but not much literature specifically concerns the integration of prevention into the local waste management. However, evidence from the literature can inform local waste management on the prevention potential; the environmental and social effects of prevention; how individuals in households can be motivated to reduce waste; and how the effects of prevention measures can be monitored. Nevertheless, knowledge is still lacking on local waste prevention, especially regarding the methods for monitoring and how local waste management systems can be designed to encourage waste reduction in the households. We end the article with recommendations for future research. The literature review can be useful for both practitioners in the waste sector and for academics seeking an overview of previous research on waste prevention. © The Author(s) 2016.

Waste container weighing data processing to create reliable information of household waste generation.

Science.gov (United States)

Korhonen, Pirjo; Kaila, Juha

2015-05-01

Household mixed waste container weighing data was processed by knowledge discovery and data mining techniques to create reliable information of household waste generation. The final data set included 27,865 weight measurements covering the whole year 2013 and it was selected from a database of Helsinki Region Environmental Services Authority, Finland. The data set contains mixed household waste arising in 6m(3) containers and it was processed identifying missing values and inconsistently low and high values as errors. The share of missing values and errors in the data set was 0.6%. This provides evidence that the waste weighing data gives reliable information of mixed waste generation at collection point level. Characteristic of mixed household waste arising at the waste collection point level is a wide variation between pickups. The seasonal variation pattern as a result of collective similarities in behaviour of households was clearly detected by smoothed medians of waste weight time series. The evaluation of the collection time series against the defined distribution range of pickup weights on the waste collection point level shows that 65% of the pickups were from collection points with optimally dimensioned container capacity and the collection points with over- and under-dimensioned container capacities were noted in 9.5% and 3.4% of all pickups, respectively. Occasional extra waste in containers occurred in 21.2% of the pickups indicating the irregular behaviour of individual households. The results of this analysis show that processing waste weighing data using knowledge discovery and data mining techniques provides trustworthy information of household waste generation and its variations. Copyright © 2015 Elsevier Ltd. All rights reserved.

Acceptable knowledge summary report for combustible/noncombustible, metallic, and HEPA filter waste resulting from 238Pu fabrication activities

International Nuclear Information System (INIS)

Rogers, P.S.Z.; Foxx, C.L.

1998-01-01

All transuranic (TRU) waste must be sufficiently characterized and certified before it is shipped to the Waste Isolation Pilot Plant (WIPP). The US Environmental Protection Agency (EPA) allows use of acceptable knowledge (AK) for waste characterization. EPA uses the term AK in its guidance document and defines AK and provides guidelines on how acceptable knowledge should be obtained and documented. This AK package has been prepared in accordance with Acceptable Knowledge Documentation (TWCP-QP-1.1-021,R.2). This report covers acceptable knowledge information for five waste streams generated at TA-55 during operations to fabricate various heat sources using feedstock 238 Pu supplied by the Savannah River Site (SRS). The 238 Pu feedstock itself does not contain quantities of RCRA-regulated constituents above regulatory threshold limits, as known from process knowledge at SRS and as confirmed by chemical analysis. No RCRA-regulated chemicals were used during 238 Pu fabrication activities at TA-55, and all 238 Pu activities were physically separated from other plutonium processing activities. Most of the waste generated from the 238 Pu fabrication activities is thus nonmixed waste, including waste streams TA-55-43, 45, and 47. The exceptions are waste streams TA-55-44, which contains discarded lead-lined rubber gloves used in the gloveboxes that contained the 238 Pu material, and TA-55-46, which may contain pieces of discarded lead. These waste streams have been denoted as mixed because of the presence of the lead-containing material

Comparison of high-solids to liquid anaerobic co-digestion of food waste and green waste

OpenAIRE

Sanati, Mehri; Chen, Xiang; Yan, Wei

2014-01-01

Co-digestion of food waste and green waste was conducted with six feedstock mixing ratios to evaluate biogas production. Increasing the food waste percentage in the feedstock resulted in an increased methane yield, while shorter retention time was achieved by increasing the green waste percentage. Food waste/green waste ratio of 40:60 was determined as preferred ratio for optimal biogas production. About 90% of methane yield was obtained after 24.5 days of digestion, with total methane yield ...

Domestic waste disposal practice and perceptions of private sector waste management in urban Accra

Science.gov (United States)

2014-01-01

Background Waste poses a threat to public health and the environment if it is not stored, collected, and disposed of properly. The perception of waste as an unwanted material with no intrinsic value has dominated attitudes towards disposal. This study investigates the domestic waste practices, waste disposal, and perceptions about waste and health in an urban community. Methods The study utilised a mixed-method approach. A cross-sectional survey questionnaire and in-depth interview were used to collect data. A total of 364 household heads were interviewed in the survey and six key informants were interviewed with the in-depth interviews. Results The results of the study revealed that 93.1% of households disposed of food debris as waste and 77.8% disposed of plastic materials as waste. The study also showed that 61.0% of the households disposed of their waste at community bins or had waste picked up at their homes by private contractors. The remaining 39.0% disposed of their waste in gutters, streets, holes and nearby bushes. Of those who paid for the services of private contractors, 62.9% were not satisfied with the services because of their cost and irregular collection. About 83% of the respondents were aware that improper waste management contributes to disease causation; most of the respondents thought that improper waste management could lead to malaria and diarrhoea. There was a general perception that children should be responsible for transporting waste from the households to dumping sites. Conclusion Proper education of the public, the provision of more communal trash bins, and the collection of waste by private contractors could help prevent exposing the public in municipalities to diseases. PMID:25005728

Methodology development for radioactive waste treatment of CDTN/BR - liquid low-level radioactive wastes

International Nuclear Information System (INIS)

Morais, Carlos Antonio de

1996-01-01

The radioactive liquid wastes generated in Nuclear Technology Development Centre (CDTN) were initially treated by precipitation/filtration and then the resulting wet solid wastes were incorporated in cement. These wastes were composed of different chemicals and different radioactivities and were generated by different sectors. The objective of the waste treatment method was to obtain minimum wet solid waste volume and decontamination and minimum operational cost. The composition of the solid wastes were taken into consideration for compatible cementation process. Approximately 5,400 litres of liquid radioactive wastes were treated by this process during 1992-1995. The volume reduction was 1/24 th and contained 20% solids. (author)

Performance test results of noninvasive characterization of Resource Conservation and Recovery Act surrogate waste by prompt gamma neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Gehrke, R.J.; Streier, G.G.

1997-03-01

During FY-96, a performance test was carried out with funding from the Mixed Waste Focus Area (MWFA) of the Department of Energy (DOE) to determine the noninvasive elemental assay capabilities of commercial companies for Resource Conservation and Recovery Act (RCRA) metals present in 8-gal drums containing surrogate waste. Commercial companies were required to be experienced in the use of prompt gamma neutron activation analysis (PGNAA) techniques and to have a prototype assay system with which to conduct the test assays. Potential participants were identified through responses to a call for proposals advertised in the Commerce Business Daily and through personal contacts. Six companies were originally identified. Two of these six were willing and able to participate in the performance test, as described in the test plan, with some subsidizing from the DOE MWFA. The tests were conducted with surrogate sludge waste because (1) a large volume of this type of waste awaits final disposition and (2) sludge tends to be somewhat homogeneous. The surrogate concentrations of the above RCRA metals ranged from {approximately} 300 ppm to {approximately} 20,000 ppm. The lower limit was chosen as an estimate of the expected sensitivity of detection required by noninvasive, pretreatment elemental assay systems to be of value for operational and compliance purposes and to still be achievable with state-of-the-art methods of analysis. The upper limit of {approximately} 20,000 ppm was chosen because it is the opinion of the author that assay above this concentration level is within current state-of-the-art methods for most RCRA constituents. This report is organized into three parts: Part 1, Test Plan to Evaluate the Technical Status of Noninvasive Elemental Assay Techniques for Hazardous Waste; Part 2, Participants` Results; and Part 3, Evaluation of and Comments on Participants` Results.

Performance test results of noninvasive characterization of Resource Conservation and Recovery Act surrogate waste by prompt gamma neutron activation analysis

International Nuclear Information System (INIS)

Gehrke, R.J.; Streier, G.G.

1997-03-01

During FY-96, a performance test was carried out with funding from the Mixed Waste Focus Area (MWFA) of the Department of Energy (DOE) to determine the noninvasive elemental assay capabilities of commercial companies for Resource Conservation and Recovery Act (RCRA) metals present in 8-gal drums containing surrogate waste. Commercial companies were required to be experienced in the use of prompt gamma neutron activation analysis (PGNAA) techniques and to have a prototype assay system with which to conduct the test assays. Potential participants were identified through responses to a call for proposals advertised in the Commerce Business Daily and through personal contacts. Six companies were originally identified. Two of these six were willing and able to participate in the performance test, as described in the test plan, with some subsidizing from the DOE MWFA. The tests were conducted with surrogate sludge waste because (1) a large volume of this type of waste awaits final disposition and (2) sludge tends to be somewhat homogeneous. The surrogate concentrations of the above RCRA metals ranged from ∼ 300 ppm to ∼ 20,000 ppm. The lower limit was chosen as an estimate of the expected sensitivity of detection required by noninvasive, pretreatment elemental assay systems to be of value for operational and compliance purposes and to still be achievable with state-of-the-art methods of analysis. The upper limit of ∼ 20,000 ppm was chosen because it is the opinion of the author that assay above this concentration level is within current state-of-the-art methods for most RCRA constituents. This report is organized into three parts: Part 1, Test Plan to Evaluate the Technical Status of Noninvasive Elemental Assay Techniques for Hazardous Waste; Part 2, Participants' Results; and Part 3, Evaluation of and Comments on Participants' Results

Radioactive Waste in Perspective

International Nuclear Information System (INIS)

2011-01-01

Large volumes of hazardous wastes are produced each year, however only a small proportion of them are radioactive. While disposal options for hazardous wastes are generally well established, some types of hazardous waste face issues similar to those for radioactive waste and also require long-term disposal arrangements. The objective of this NEA study is to put the management of radioactive waste into perspective, firstly by contrasting features of radioactive and hazardous wastes, together with their management policies and strategies, and secondly by examining the specific case of the wastes resulting from carbon capture and storage of fossil fuels. The study seeks to give policy makers and interested stakeholders a broad overview of the similarities and differences between radioactive and hazardous wastes and their management strategies. Contents: - Foreword; - Key Points for Policy Makers; - Executive Summary; - Introduction; - Theme 1 - Radioactive and Hazardous Wastes in Perspective; - Theme 2 - The Outlook for Wastes Arising from Coal and from Nuclear Power Generation; - Risk, Perceived Risk and Public Attitudes; - Concluding Discussion and Lessons Learnt; - Strategic Issues for Radioactive Waste; - Strategic Issues for Hazardous Waste; - Case Studies - The Management of Coal Ash, CO 2 and Mercury as Wastes; - Risk and Perceived Risk; - List of Participants; - List of Abbreviations. (authors)

Results of HWVP transuranic process waste treatment laboratory and pilot-scale filtration tests using specially ground zeolite

International Nuclear Information System (INIS)

Eakin, D.E.

1996-03-01

Process waste streams from the Hanford Waste Vitrification Plant (HWVP) may require treatment for cesium, strontium, and transuranic (TRU) element removal in order to meet criteria for incorporation in grout. The approach planned for cesium and strontium removal is ion exchange using a zeolite exchanger followed by filtration. Filtration using a pneumatic hydropulse filter is planned to remove TRU elements which are associated with process solids and to also remove zeolite bearing the cesium and strontium. The solids removed during filtration are recycled to the melter feed system to be incorporated into the HWVP glass product. Fluor Daniel, Inc., the architect-engineering firm for HWVP, recommended a Pneumatic Hydropulse (PHP) filter manufactured by Mott Metallurgical Corporation for use in the HWVP. The primary waste streams considered for application of zeolite contact and filtration are melter off-gas condensate from the submerged bed scrubber (SBS), and equipment decontamination solutions from the Decontamination Waste Treatment Tank (DWTT). Other waste streams could be treated depending on TRU element and radionuclide content. Laboratory and pilot-scale filtration tests were conducted to provide a preliminary assessment of the adequacy of the recommended filter for application to HWVP waste treatment

Radionuclide Incorporation in Secondary Crystalline Minerals Resulting from Chemical Weathering of Selected Waste Glasses: Progress Report: Task kd.5b

International Nuclear Information System (INIS)

Mattigod, Shas V.; Serne, R. Jeffrey; Legore, Virginia L.; Parker, Kent E.; Orr, Robert D.; McCready, David E.; )

2003-01-01

Experiments were conducted by Pacific Northwest National Laboratory to evaluate potential incorporation of radionuclides in secondary mineral phases that form from weathering vitrified nuclear waste glasses. These experiments were conducted as part of the Immobilized Low-Activity Waste-Performance Assessment (ILAW-PA) to generate data on radionuclide mobilization and transport in a near-field environment of disposed vitrified wastes. The results of these experiments demonstrated that radionuclide sequestration can be significantly enhanced by promoting the formation of cage structured minerals such as sodalite from weathering glasses. These results have important implications regarding radionuclide sequestration/mobilization aspects that are not currently accounted for in the ILAW PA. Additional studies are required to confirm the results and to develop an improved understanding of the mechanisms of sequestration of radionuclides into the secondary and tertiary weathering products o f the ILAW glass to help refine how contaminants are released from the near-field disposal region out into the accessible environment. Of particular interest is to determine whether the contaminants remain sequestered in the glass weathering products for hundreds to thousands of years. If the sequestration can be shown to continue for long periods, another immobilization process can be added to the PA analysis and predicted risks should be lower than past predictions

Incineration of a typical LWR combustible waste and analysis of the resulting ash

International Nuclear Information System (INIS)

Treat, R.L.; Lokken, R.O.; Schliebe, M.J.

1983-05-01

In this study 4540 kg (10,000 lb) of simulated nuclear power plant combustion wastes were burned in a controlled-air incinerators. The purpose of this work was to generate ashes suitable for solidification, the products of which will be analyzed to determine if they are suitable for disposal. Two different types of waste were burned: resin and simulated crud, and general trash (paper, plastics, wood, rubber, and cloth). Volume-reduction ratios (unburned waste: ash) were 13:1 and 22:1, respectively. Approximately 20% of the ash was lost due to adherence to incinerator walls and entrainment in the off-gas stream. Losses of the volatile species cesium and iodine were 79% and 100%, respectively. The ashes were not hygroscopic, but they exhibited a pH of 4.6 to 5.0 when water was added. Corrosion of mild steel drums would occur within this pH range. The ashes contained a significant quantity of clinkers haveing lengths as great as 20 cm (8 in.). Most of the clinkers were fully incinerated and easy to crush, suggesting that standard comminuting equipment should be effective in reducing the size of clinkers to allow their solidification with the fine ashes

Liquid Secondary Waste Grout Formulation and Waste Form Qualification

Energy Technology Data Exchange (ETDEWEB)

Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williams, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle M. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-05-23

This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the ⁹⁹Tc desorption K_d (distribution coefficient) values for ⁹⁹Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of ⁹⁹Tc(IV)-bearing solid phases for ⁹⁹Tc transport in reducing conditions to support the IDF PA.

Data analytics approach to create waste generation profiles for waste management and collection.

Science.gov (United States)

Niska, Harri; Serkkola, Ari

2018-04-30

Extensive monitoring data on waste generation is increasingly collected in order to implement cost-efficient and sustainable waste management operations. In addition, geospatial data from different registries of the society are opening for free usage. Novel data analytics approaches can be built on the top of the data to produce more detailed, and in-time waste generation information for the basis of waste management and collection. In this paper, a data-based approach based on the self-organizing map (SOM) and the k-means algorithm is developed for creating a set of waste generation type profiles. The approach is demonstrated using the extensive container-level waste weighting data collected in the metropolitan area of Helsinki, Finland. The results obtained highlight the potential of advanced data analytic approaches in producing more detailed waste generation information e.g. for the basis of tailored feedback services for waste producers and the planning and optimization of waste collection and recycling. Copyright © 2018 Elsevier Ltd. All rights reserved.

Dynamics of radioactive waste generation

International Nuclear Information System (INIS)

Dogaru, Daniela; Virtopeanu, Cornelia; Ivan, Alexandrina

2008-01-01

In Romania there are in operation three facilities licensed for collection, treatment and storage of radioactive waste resulted from industry, research, medicine, and agriculture, named institutional radioactive waste. The repository, which is of near surface type, is designed for disposing institutional radioactive waste. The institutional radioactive wastes generated are allowed to be disposed into repository according to the waste acceptance criteria, defined for the disposal facility. The radioactive wastes which are not allowed for disposal are stored on the site of each facility which is special authorised for this. The paper describes the dynamics of generation of institutional waste in Romania, both for radioactive waste which are allowed to be disposed into repository and for radioactive waste which are not allowed to be disposed of. (authors)

Socio-economic impact of improper hospital waste management on waste disposal employees

International Nuclear Information System (INIS)

Khan, M.R.; Raza, Z. L.

2011-01-01

Background: Improper disposal of hospital waste results in spread of disease to the community and its handlers. Objectives: To study the socio-economic impact of inappropriate disposal of hospital waste on the health of the waste disposal staff. Materials and Methods: Interviews were conducted from 50 hospital waste collectors of Lahore and using a pre-structured questionnaire, the information was filled. The data were statistically analyzed for frequencies, and cross tabulation. Results: The improper disposal of hospital waste lead to disease in 45 hospital waste collectors. Eighteen waste collectors were infected with respiratory diseases,14 with skin infection, 7 with tuberculosis and 6 with hepatitis. Only 8 workers were provided with special clothes by the hospital management. The chances of getting infection was high in those who were not provided with special clothes like, gowns, gloves and shoes as compared to those who were provided with these.The total cost of recovery for these diseases also varied with an amount of Rs. 68,340 for the treatment of hepatitis, Rs. 3,150 for tuberculosis, Rs. 1,500 for respiratory diseases and Rs. 1,000 for skin infection. Only 12 workers were given a small remuneration ranging from Rs.100-400 per month as compensation from the hospital administration. Conclusions: Use of protective clothing by the hospital waste disposal collectors can significantly reduce their exposure to the diseases. Policy message: Provision of clothing and gloves to the waste disposal collectors, would help significantly in reducing diseases like tuberculosis, hepatitis, respiratory diseases and skin infection. (author)

Ferrocyanide Safety Program: Analysis of postulated energetic reactions and resultant aerosol generation in Hanford Site Waste Tanks

International Nuclear Information System (INIS)

Postma, A.K.; Dickinson, D.R.

1995-09-01

This report reviews work done to estimate the possible consequences of postulated energetic reactions in ferrocyanide waste stored in underground tanks at the Hanford Site. The issue of explosive reactions was raised in the 1987 Environmental Impact Statement (EIS), where a detonation-like explosion was postulated for the purpose of defining an upper bound on dose consequences for various disposal options. A review of the explosion scenario by the General Accounting Office (GAO) indicated that the aerosol generation and consequent radioactive doses projected for the explosion postulated in the EIS were understated by one to two orders of magnitude. The US DOE has sponsored an extensive study of the hazard posed by uncontrolled exothermic reactions in ferrocyanide waste, and results obtained during the past three years have allowed this hazard to be more realistically assessed. The objective of this report is to summarize the improved knowledge base that now indicates that explosive or vigorous chemical reactions are not credible in the ferrocyanide waste stored in underground tanks. This improved understanding supports the decision not to proceed with further analyses or predictions of the consequences of such an event or with aerosol tests in support of such predictions. 53 refs., 2 tabs

Intelligent Information System for Waste Management; Jaetehuollon aelykaes tietojaerjestelmae iWaste

Energy Technology Data Exchange (ETDEWEB)

Mustonen, T. [Kuopio Univ. (Finland)

2003-07-01

'iWaste' is a project for developing and testing intelligent computational methods for more comprehensive waste management. Important issues are automated reporting, optimisation of waste collection, forecasting of waste formation, data handling of waste disposal sites and simulation and modelling of regional waste management. The main objective of the project is to identify and analyse known sources of information and to link them to the existing information processing systems in the field of waste management. Additionally, the goal is to identify and test functional elements that could be developed further to software products and services. The results of the project can be categorized into three sectors. Firstly, the guidelines for a comprehensive information system in waste management will be created. This includes the requirement specifications of different parties, definitions for the data exchange interfaces and an architectural plan for software products capable of co-operative processing. Secondly, the central parts of the intelligent information system will be piloted using the research database collected in the early stage of the project. The main topics investigated are data quality, the use of Geographical Information Systems (GIS), automated reporting, optimisation of waste collection and forecasting of waste formation. Additionally, the pilot information system can be utilized in derivative projects to speed up the starting phases of them. This makes it possible to create persistent development of waste management information systems both academically and commercially. (orig.)

Analysis of low-level wastes. Review of hazardous waste regulations and identification of radioactive mixed wastes. Final report

International Nuclear Information System (INIS)

Bowerman, B.S.; Kempf, C.R.; MacKenzie, D.R.; Siskind, B.; Piciulo, P.L.

1985-12-01

Regulations governing the management and disposal of hazardous wastes have been promulgated by the US Environmental Protection Agency under authority of the Resource Conservation and Recovery Act. These were reviewed and compared with the available information on the properties and characteristics of low-level radioactive wastes (LLW). In addition, a survey was carried out to establish a data base on the nature and composition of LLW in order to determine whether some LLW streams could also be considered hazardous as defined in 40 CFR Part 261. For the survey, an attempt was made to obtain data on the greatest volume of LLW; hence, as many large LLW generators as possible were contacted. The list of 238 generators contacted was based on information obtained from NRC and other sources. The data base was compiled from completed questionnaires which were returned by 97 reactor and non-reactor facilities. The waste volumes reported by these respondents corresponded to approximately 29% of all LLW disposed of in 1984. The analysis of the survey results indicated that three broad categories of LLW may be radioactive mixed wastes. They include: waste containing organic liquids, disposed of by all types of generators; wastes containing lead metal, i.e., discarded shielding or lead containers; wastes containing chromates, i.e., nuclear power plant process wastes where chromates are used as corrosion inhibitors. Certain wastes, specific to particular generators, were identified as potential mixed wastes as well. 8 figs., 48 tabs

Solidification of intermediate level liquid waste - ILLW, CEMEX waste form qualification

International Nuclear Information System (INIS)

D'Andrea, V.; Guerra, M.; Pancotti, F.; Maio, V.

2015-01-01

In the Sogin EUREX Facility about 125 m 3 of intermediate level radioactive waste and about 113 m 3 of low level radioactive waste, produced during the re-processing of MTR and CANDU fuel, are stored. Solidification of these wastes is planned in order to fulfill the specific requirements established by the Safety Authority, taking into account the criteria set up in a Technical Guide on the issue of radioactive waste management. The design of a cementation plant (CEMEX) of all liquid radioactive wastes is currently ongoing. The process requires that the liquid waste is neutralized with NaOH (NaOH 19 M) and metered into 440 liter drum together with the cement, while the mixture is stirred by a lost paddle ('in drum mixing process'). The qualification of the Waste Form consists of all the activities demonstrating that the final cemented product has the minimum requirements (mechanical, chemical and physical characteristics) compliant with all the subsequent management phases: long-term interim storage, transport and long-term disposal of the waste. All tests performed to qualify the conditioning process for immobilizing first extraction cycle (MTR and CANDU) and second extraction cycle liquid wastes, gave results in compliance with the minimum requirements established for disposal

Waste gas combustion in a Hanford radioactive waste tank

International Nuclear Information System (INIS)

Travis, J.R.; Fujita, R.K.; Spore, J.W.

1994-01-01

It has been observed that a high-level radioactive waste tank generates quantities of hydrogen, ammonia, nitrous oxide, and nitrogen that are potentially well within flammability limits. These gases are produced from chemical and nuclear decay reactions in a slurry of radioactive waste materials. Significant amounts of combustible and reactant gases accumulate in the waste over a 110- to 120-d period. The slurry becomes Taylor unstable owing to the buoyancy of the gases trapped in a matrix of sodium nitrate and nitrite salts. As the contents of the tank roll over, the generated waste gases rupture through the waste material surface, allowing the gases to be transported and mixed with air in the cover-gas space in the dome of the tank. An ignition source is postulated in the dome space where the waste gases combust in the presence of air resulting in pressure and temperature loadings on the double-walled waste tank. This analysis is conducted with hydrogen mixing studies HMS, a three-dimensional, time-dependent fluid dynamics code coupled with finite-rate chemical kinetics. The waste tank has a ventilation system designed to maintain a slight negative gage pressure during normal operation. We modeled the ventilation system with the transient reactor analysis code (TRAC), and we coupled these two best-estimate accident analysis computer codes to model the ventilation system response to pressures and temperatures generated by the hydrogen and ammonia combustion

Does industrial waste taxation contribute to reduction of landfilled waste? Dynamic panel analysis considering industrial waste category in Japan.

Science.gov (United States)

Sasao, Toshiaki

2014-11-01

Waste taxes, such as landfill and incineration taxes, have emerged as a popular option in developed countries to promote the 3Rs (reduce, reuse, and recycle). However, few studies have examined the effectiveness of waste taxes. In addition, quite a few studies have considered both dynamic relationships among dependent variables and unobserved individual heterogeneity among the jurisdictions. If dependent variables are persistent, omitted variables cause a bias, or common characteristics exist across the jurisdictions that have introduced waste taxes, the standard fixed effects model may lead to biased estimation results and misunderstood causal relationships. In addition, most existing studies have examined waste in terms of total amounts rather than by categories. Even if significant reductions in total waste amounts are not observed, some reduction within each category may, nevertheless, become evident. Therefore, this study analyzes the effects of industrial waste taxation on quantities of waste in landfill in Japan by applying the bias-corrected least-squares dummy variable (LSDVC) estimators; the general method of moments (difference GMM); and the system GMM. In addition, the study investigates effect differences attributable to industrial waste categories and taxation types. This paper shows that industrial waste taxes in Japan have minimal, significant effects on the reduction of final disposal amounts thus far, considering dynamic relationships and waste categories. Copyright © 2014 Elsevier Ltd. All rights reserved.

Treatment of mercury containing waste

Science.gov (United States)

Kalb, Paul D.; Melamed, Dan; Patel, Bhavesh R; Fuhrmann, Mark

2002-01-01

A process is provided for the treatment of mercury containing waste in a single reaction vessel which includes a) stabilizing the waste with sulfur polymer cement under an inert atmosphere to form a resulting mixture and b) encapsulating the resulting mixture by heating the mixture to form a molten product and casting the molten product as a monolithic final waste form. Additional sulfur polymer cement can be added in the encapsulation step if needed, and a stabilizing additive can be added in the process to improve the leaching properties of the waste form.

Municipal solid waste composition determination supporting the integrated solid waste management system in the island of Crete

International Nuclear Information System (INIS)

Gidarakos, E.; Havas, G.; Ntzamilis, P.

2006-01-01

A one-year survey was conducted in the greater region of Crete (located at the lower region of the Aegean Sea) for the purpose of identifying waste composition (including chemical and physical characterization), as well as any seasonal variation. The investigation was carried out repeatedly at seven landfills and one transfer station in Crete, in four phases. Each sampling phase corresponded to a season (autumn, winter, spring, summer). ASTM D5231-92(2003) standard method and RCRA Waste Sampling Draft Technical Guidance were used. Hand sorting was used for classifying the collected wastes into the following categories: plastics, paper, metals, aluminium, leather-wood-textiles-rubbers, organic wastes, non-combustibles and miscellaneous. Further analysis included proximate and ultimate analysis of combustible materials. Metals such as lead, cadmium and mercury were also investigated. The results show that there has been a significant decrease of organic wastes during the last decade due to the increase of packaging materials, as a result of a change in consumption patterns. Three main waste categories were determined: organic wastes, paper and plastics, which combined represent 76% of the total waste in Crete. Furthermore, a high fraction of glass and a seasonal variation of aluminium indicate a strong correlation of waste composition with certain human activities, such as tourism. There is also a variation between the municipal solid waste (MSW) composition in the region of Crete (2003-2004) and MSW composition suggested in the National Solid Waste Planning (2000) [National Solid Waste Planning, 2000. Completion and particularization of Common Ministerial Act 113944//1944/1997: National Solid Waste Planning, June 2000]. The results of this survey are to be utilized by the regional solid waste authorities in order to establish an integrated waste treatment site, capable of fulfilling the regional waste management demands

Municipal solid waste composition determination supporting the integrated solid waste management system in the island of Crete.

Science.gov (United States)

Gidarakos, E; Havas, G; Ntzamilis, P

2006-01-01

A one-year survey was conducted in the greater region of Crete (located at the lower region of the Aegean Sea) for the purpose of identifying waste composition (including chemical and physical characterization), as well as any seasonal variation. The investigation was carried out repeatedly at seven landfills and one transfer station in Crete, in four phases. Each sampling phase corresponded to a season (autumn, winter, spring, summer). ASTM D5231-92(2003) standard method and RCRA Waste Sampling Draft Technical Guidance were used. Hand sorting was used for classifying the collected wastes into the following categories: plastics, paper, metals, aluminium, leather-wood-textiles-rubbers, organic wastes, non-combustibles and miscellaneous. Further analysis included proximate and ultimate analysis of combustible materials. Metals such as lead, cadmium and mercury were also investigated. The results show that there has been a significant decrease of organic wastes during the last decade due to the increase of packaging materials, as a result of a change in consumption patterns. Three main waste categories were determined: organic wastes, paper and plastics, which combined represent 76% of the total waste in Crete. Furthermore, a high fraction of glass and a seasonal variation of aluminium indicate a strong correlation of waste composition with certain human activities, such as tourism. There is also a variation between the municipal solid waste (MSW) composition in the region of Crete (2003-2004) and MSW composition suggested in the National Solid Waste Planning (2000) [National Solid Waste Planning, 2000. Completion and particularization of Common Ministerial Act 113944//1944/1997: National Solid Waste Planning, June 2000]. The results of this survey are to be utilized by the regional solid waste authorities in order to establish an integrated waste treatment site, capable of fulfilling the regional waste management demands.

The characterization of cement waste form for final disposal of decommissioned concrete waste

International Nuclear Information System (INIS)

Lee, K.W.; Lee, Y.J.; Hwang, D.S.; Moon, J.K.

2015-01-01

Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. In addition, 83 drums of 200 l, and 41 containers of 4 m 3 of concrete waste were generated. Conditioning of concrete waste is needed for final disposal. Concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled into a void space after concrete rubble pre-placement into 200 l drums. Thus, this research developed an optimizing mixing ratio of concrete waste, water, and cement, and evaluated the characteristics of a cement waste form to meet the requirements specified in the disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, and thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10 as the optimized mixing ratio. In addition, the compressive strength of cement waste form was satisfied, including fine powder up to a maximum 40 wt% in concrete debris waste of about 75%. (authors)

Comparison of high-solids to liquid anaerobic co-digestion of food waste and green waste.

Science.gov (United States)

Chen, Xiang; Yan, Wei; Sheng, Kuichuan; Sanati, Mehri

2014-02-01

Co-digestion of food waste and green waste was conducted with six feedstock mixing ratios to evaluate biogas production. Increasing the food waste percentage in the feedstock resulted in an increased methane yield, while shorter retention time was achieved by increasing the green waste percentage. Food waste/green waste ratio of 40:60 was determined as preferred ratio for optimal biogas production. About 90% of methane yield was obtained after 24.5 days of digestion, with total methane yield of 272.1 mL/g VS. Based the preferred ratio, effect of total solids (TS) content on co-digestion of food waste and green waste was evaluated over a TS range of 5-25%. Results showed that methane yields from high-solids anaerobic digestion (15-20% TS) were higher than the output of liquid anaerobic digestion (5-10% TS), while methanogenesis was inhibited by further increasing the TS content to 25%. The inhibition may be caused by organic overloading and excess ammonia. Copyright © 2013 Elsevier Ltd. All rights reserved.

Radon emission from uranium mining waste rock dumps and resulting radon immission; Radonemissionsverhalten von Halden des Uranbergbaus und daraus resultierende Radonemissionen

Energy Technology Data Exchange (ETDEWEB)

Regner, J.; Hinz, W.; Schmidt, P. [Wismut GmbH, Chemnitz (Germany)

2016-07-01

Since more than 20 years, Wismut GmbH has been investigating the radon situation at uranium mining waste rock dumps. In the present paper the results of 19 complex studies at uranium mining dumps in the Erzgebirge (Ore Mountains) are reported. Although the mean specific activity of Ra-226 of the waste rock material was on a rather low level of about 0.5 Bq/g, the mean radon concentration in free atmosphere at the public exposure sites in the immediate vicinity of the dumps reached a value of about 1000 Bq/m{sup 3} for a half-year exposition and of about 600 Bq/m{sup 3} for a one-year exposition. Certain geometries and structures of waste rock dumps and the occurrence of convective airflows in the dumps are main reasons for the high radon emission despite of the relatively low specific Ra-226 activity. A case study for two buildings directly on the top of a waste rock dump in the town Johanngeorgenstadt is presented. The hypothetical interpolation of the results for Ra-226-activity to a value below the threshold value of 0.2 Bq/g leads to the assumption that problematic radon situations may also occur outside the areas of legacies of uranium mining. Considering the aspects mentioned, a clearance level for NORM of 1 Bq/g is questionable.

Managing soil moisture on waste burial sites

International Nuclear Information System (INIS)

Anderson, J.E.; Ratzlaff, T.D.

1991-11-01

Shallow land burial is a common method of disposing of industrial, municipal, and low-level radioactive waste. The exclusion of water from buried wastes is a primary objective in designing and managing waste disposal sites. If wastes are not adequately isolated, water from precipitation may move through the landfill cover and into the wastes. The presence of water in the waste zone may promote the growth of plant roots to that depth and result in the transport of toxic materials to above-ground foliage. Furthermore, percolation of water through the waste zone may transport contaminants into ground water. This report presents results from a field study designed to assess the the potential for using vegetation to deplete soil moisture and prevent water from reaching buried wastes at the Idaho National Engineering Laboratory (INEL). Our results show that this approach may provide an economical means of limiting the intrusion of water on waste sites

Krsko NPP radioactive waste characteristics

International Nuclear Information System (INIS)

Skanata, D.; Kroselj, V.; Jankovic, M.

2007-01-01

In May 2005 Krsko NPP initiated the Radioactive Waste Characterization Project and commissioned its realization to the consulting company Enconet International, Zagreb. The Agency for Radwaste Management was invited to participate on the Project. The Project was successfully closed out in August 2006. The main Project goal consisted of systematization the existing and gathering the missing radiological, chemical, physical, mechanical, thermal and biological information and data on radioactive waste. In a general perspective, the Project may also be considered as a part of broader scope of activities to support state efforts to find a disposal solution for radioactive waste in Slovenia. The operational low and intermediate level radioactive waste has been structured into 6 waste streams that contain evaporator concentrates and tank sludges, spent ion resins, spent filters, compressible and non-compressible waste as well as specific waste. For each of mentioned waste streams, process schemes have been developed including raw waste, treatment and conditioning technologies, waste forms, containers and waste packages. In the paper the main results of the Characterization Project will be briefly described. The results will indicate that there are 17 different types of raw waste that have been processed by applying 9 treatment/conditioning technologies. By this way 18 different waste forms have been produced and stored into 3 types of containers. Within each type of container several combinations should be distinguished. Considering all of this, there are 34 different types of waste packages altogether that are currently stored in the Solid Radwaste Storage Facility at the Krsko NPP site. Because of these findings a new identification system has been recommended and consequently the improvement of the existing database on radioactive waste has been proposed. The potential areas of further in depth characterization are indicated. In the paper a brief description on the

Headspace vapor characterization of Hanford Waste Tank 241-U-112: Results from samples collected on 7/09/96

International Nuclear Information System (INIS)

Evans, J.C.; Pool, K.H.; Thomas, B.L.; Olsen, K.B.; Fruchter, J.S.; Silvers, K.L.

1997-01-01

This report describes the analytical results of vapor samples taken from the headspace of the waste storage tank 241-U-112 at the Hanford Site in Washington State. The results described in this report were obtained to characterize the vapors present in the tank headspace and to support safety evaluations and tank farm operations. The results include air concentrations of selected inorganic and organic analytes and grouped compounds from samples obtained by Westinghouse Hanford Company

The results of an ecological risk assessment screening at the Idaho National Engineering's waste area group 2

International Nuclear Information System (INIS)

VanHorn, R.

1995-01-01

The Idaho National Engineering Laboratory (INEL) is a Department of Energy (DOE) facility located in southeastern Idaho and occupies approximately 890 square miles on the northwestern portion of the eastern Snake River Plain. INEL has been devoted to nuclear energy research and related activities since its establishment in 1949. In the process of fulfilling this mission, wastes were generated, including radioactive and hazardous materials. Most materials were effectively stored or disposed of, however, some release of contaminants to the environment has occurred. For this reason, the INEL was listed by the US environmental Protection Agency on the National Priorities List (NPL), in November, 1989. This report describes the results of an ecological risk assessment performed for the Waste Area Groups 2 (WAG 2) at the INEL. It also summarizes the performance of screening level ecological risk assessments (SLERA)

Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

Energy Technology Data Exchange (ETDEWEB)

Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

2011-09-12

The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

Grout treatment facility dangerous waste permit application

International Nuclear Information System (INIS)

1992-07-01

The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the low-level liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Hanford Site Maps, road evaluation for the grout treatment facility, Department of Ecology certificate of non-designation for centralia fly ash, double-shell tank waste compositional modeling, laboratory analysis reports for double-shell tank waste, stored in tanks 241-AN-103, 241-AN-106, and 241-AW-101, grout vault heat transfer results for M-106 grout formulation, test results for extraction procedure toxicity testing, test results for toxicity testing of double-shell tank grout, pilot-scale grout production test with a simulated low-level waste, characterization of simulated low-level waste grout produced in a pilot-scale test, description of the procedure for sampling nonaging waste storage tanks, description of laboratory procedures, grout campaign waste composition verification, variability in properties of grouted phosphate/sulfate N-reactor waste, engineering drawings, description of operating procedures, equipment list--transportable grout equipment, grout treatment facility--tank integrity assessment plan, long-term effects of waste solutions on concrete and reinforcing steel, vendor information, grout disposal facilities construction quality assurance plan, and flexible membrane liner/waste compatibility test results

Incineration of radioactive waste

International Nuclear Information System (INIS)

Eid, C.

1985-01-01

The incineration process currently seems the most appropriate way to solve the problems encountered by the increasing quantities of low and medium active waste from nuclear power generation waste. Although a large number of incinerators operate in the industry, there is still scope for the improvement of safety, throughput capacity and reduction of secondary waste. This seminar intends to give opportunity to scientists working on the different aspects of incineration to present their most salient results and to discuss the possibilities of making headway in the management of LL/ML radioactive waste. These proceedings include 17 contributions ranging over the subjects: incineration of solid β-γ wastes; incineration of other radwastes; measurement and control of wastes; off-gas filtration and release. (orig./G.J.P.)

Factors affecting the rural domestic waste generation

Directory of Open Access Journals (Sweden)

A.R. Darban Astane

2017-12-01

Full Text Available The current study was carried out to evaluate the quantity and quality of rural domestic waste generation and to identify the factors affecting it in rural areas of Khodabandeh county in Zanjan Province, Iran. Waste samplings consisted of 318 rural households in 11 villages. In order to evaluate the quality and quantity of the rural domestic waste, waste production was classified into 12 groups and 2 main groups of organic waste and solid waste. Moreover, kriging interpolation technique in ARC-GIS software was used to evaluate the spatial distribution of the generated domestic waste and ultimately multiple regression analysis was used to evaluate the factors affecting the generation of domestic waste. The results of this study showed that the average waste generated by each person was 0.588 kilograms per day. with the share of organic waste generated by each person being 0.409 kilograms per day and the share of solid waste generated by each person being 0.179 kilograms per day. The results from spatial distribution of waste generation showed a certain pattern in three groups and a higher rate of waste generation in the northern and northwestern parts, especially in the subdistrict. The results of multiple regression analysis showed that the households’ income, assets, age, and personal attitude are respectively the most important variables affecting waste generation. The housholds’ attitude and indigenous knowledge on efficient use of materials are also the key factors which can help reducing waste generation.

Waste to energy – key element for sustainable waste management

Energy Technology Data Exchange (ETDEWEB)

Brunner, Paul H., E-mail: paul.h.brunner@tuwien.ac.at; Rechberger, Helmut

2015-03-15

Highlights: • First paper on the importance of incineration from a urban metabolism point of view. • Proves that incineration is necessary for sustainable waste management. • Historical and technical overview of 100 years development of MSW incineration. - Abstract: Human activities inevitably result in wastes. The higher the material turnover, and the more complex and divers the materials produced, the more challenging it is for waste management to reach the goals of “protection of men and environment” and “resource conservation”. Waste incineration, introduced originally for volume reduction and hygienic reasons, went through a long and intense development. Together with prevention and recycling measures, waste to energy (WTE) facilities contribute significantly to reaching the goals of waste management. Sophisticated air pollution control (APC) devices ensure that emissions are environmentally safe. Incinerators are crucial and unique for the complete destruction of hazardous organic materials, to reduce risks due to pathogenic microorganisms and viruses, and for concentrating valuable as well as toxic metals in certain fractions. Bottom ash and APC residues have become new sources of secondary metals, hence incineration has become a materials recycling facility, too. WTE plants are supporting decisions about waste and environmental management: They can routinely and cost effectively supply information about chemical waste composition as well as about the ratio of biogenic to fossil carbon in MSW and off-gas.

Waste to energy – key element for sustainable waste management

International Nuclear Information System (INIS)

Brunner, Paul H.; Rechberger, Helmut

2015-01-01

Highlights: • First paper on the importance of incineration from a urban metabolism point of view. • Proves that incineration is necessary for sustainable waste management. • Historical and technical overview of 100 years development of MSW incineration. - Abstract: Human activities inevitably result in wastes. The higher the material turnover, and the more complex and divers the materials produced, the more challenging it is for waste management to reach the goals of “protection of men and environment” and “resource conservation”. Waste incineration, introduced originally for volume reduction and hygienic reasons, went through a long and intense development. Together with prevention and recycling measures, waste to energy (WTE) facilities contribute significantly to reaching the goals of waste management. Sophisticated air pollution control (APC) devices ensure that emissions are environmentally safe. Incinerators are crucial and unique for the complete destruction of hazardous organic materials, to reduce risks due to pathogenic microorganisms and viruses, and for concentrating valuable as well as toxic metals in certain fractions. Bottom ash and APC residues have become new sources of secondary metals, hence incineration has become a materials recycling facility, too. WTE plants are supporting decisions about waste and environmental management: They can routinely and cost effectively supply information about chemical waste composition as well as about the ratio of biogenic to fossil carbon in MSW and off-gas

Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

International Nuclear Information System (INIS)

Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.

2011-01-01

To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that

Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

Energy Technology Data Exchange (ETDEWEB)

Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.

2011-08-12

To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that

Results of Waste Transfer and Back-Dilution in Tanks 241-SY-101 and 241-SY-102

International Nuclear Information System (INIS)

Mahoney, L.A.; Antoniak, Z.I.; Barton, W.B.; Conner, J.M.; Kirch, N.W.; Stewart, C.W.; Wells, B.E.

2000-01-01

This report chronicles the process of remediation of the flammable gas hazard in Tank 241-SY-101 (SY-101) by waste transfer and back-dilution from December 18, 1999 through April 2, 2000. A brief history is given of the development of the flammable gas retention and release hazard in this tank, and the transfer and dilution systems are outlined. A detailed narrative of each of the three transfer and dilution campaigns is given to provide structure for the balance of the report. Details of the behavior of specific data are then described, including the effect of transfer and dilution on the waste levels in Tanks SY-101 and SY-102, data from strain gauges on equipment suspended from the tank dome, changes in waste configuration as inferred from neutron and gamma logs, headspace gas concentrations, waste temperatures, and the mixerpump operating performance. Operating data and performance of the transfer pump in SY-101 are also discussed

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

Non-radiological contrast agents (MRI)

International Nuclear Information System (INIS)

Bonnemain, B.; Lautrou, J.; Meyer, D.; Doucet, D.

1987-01-01

Over the past few years, extensive research has been carried out in an attempt to develop contrast agents that could help improve both the performance (acquisition times) and the diagnostic efficacy of Magnetic Resonance Imaging (MRI) techniques. On the basis of physicochemical and pharmacological criteria discussed in this presentation, a few efficacious, well-tolerated compounds could be developed. Two of them, the gadolinium complexes Gd-DOTA and Gd-DTPA, are currently being tried in man. This first generation of contrast agents, which are aspecific markers of the intravascular space, has been shown to have good diagnostic potential in conventional MRI procedures. The diagnostic contribution of these contrast agents will probably be a most essential factor in new MRI techniques using low field strengh or fast imaging sequences [fr

Waste tank vapor project: Vapor space characterization of waste tank 241-BY-104: Results from samples collected on June 24, 1994

International Nuclear Information System (INIS)

Clauss, T.W.; Ligotke, M.W.; McVeety, B.D.; Pool, K.H.; Lucke, R.B.; Fruchter, J.S.; Goheen, S.C.

1994-11-01

This report describes results of the analyses of tank-headspace samples taken from Hanford waste Tank 241-BY-104 (referred to as Tank BY-104) on June 24, 1994. The Pacific Northwest Laboratory (PNL) contracted with Westinghouse Hanford Company (WHC) to provide sampling devices and analyze inorganic and organic samples collected from the tank headspace. The sample job was designated S4019 and was performed by WHC on June 24, 1994 using the vapor sampling system (VSS). The results of the analyses are expected to be used in the determination of safety and toxicological issues related to the tank-headspace gas as described in the WHC report entitled Data Quality Objectives for Generic In-Tank Health and Safety Vapor Issue Resolution, WHC-SD-WM-DQO-002, Rev. 0. Sampling devices, including 16 sorbent trains (for inorganic analyses), and 5 SUMMA trademark canisters (for organic analyses), were supplied to the WHC sampling staff on June 20, 1994. Samples were taken (by WHC) on June 24. The samples were returned from the field on June 27. The inorganic samples delivered to PNL on chain-of-custody (COC) 006893 included 16 sorbent trains as described in Tables 2.2, 2.3, and 2.4. Additional inorganic blank spikes were obtained from related sample jobs. SUMMA trademark samples delivered to PNL on COC 006896 included one ambient air sample, one ambient-air sample through the sampling system, and three tank-headspace SUMMA trademark canister samples. The samples were inspected upon delivery to the 326/23B laboratory and logged into PNL laboratory record book 55408. Custody of the sorbent trains was transferred to PNL personnel performing the inorganic analysis and stored at refrigerated (≤10 degrees C) temperature until the time of analysis. Access to the 326/23B laboratory is limited to PNL personnel working on the waste-tank safety program

Washing of waste prior to landfilling.

Science.gov (United States)

Cossu, Raffaello; Lai, Tiziana

2012-05-01

The main impact produced by landfills is represented by the release of leachate emissions. Waste washing treatment has been investigated to evaluate its efficiency in reducing the waste leaching fraction prior to landfilling. The results of laboratory-scale washing tests applied to several significant residues from integrated management of solid waste are presented in this study, specifically: non-recyclable plastics from source separation, mechanical-biological treated municipal solid waste and a special waste, automotive shredded residues. Results obtained demonstrate that washing treatment contributes towards combating the environmental impacts of raw wastes. Accordingly, a leachate production model was applied, leading to the consideration that the concentrations of chemical oxygen demand (COD) and total Kjeldahl nitrogen (TKN), parameters of fundamental importance in the characterization of landfill leachate, from a landfill containing washed wastes, are comparable to those that would only be reached between 90 and 220years later in the presence of raw wastes. The findings obtained demonstrated that washing of waste may represent an effective means of reducing the leachable fraction resulting in a consequent decrease in landfill emissions. Further studies on pilot scale are needed to assess the potential for full-scale application of this treatment. Copyright © 2012 Elsevier Ltd. All rights reserved.

Waste Management Process Improvement Project

International Nuclear Information System (INIS)

Atwood, J.; Borden, G.; Rangel, G. R.

2002-01-01

The Bechtel Hanford-led Environmental Restoration Contractor team's Waste Management Process Improvement Project is working diligently with the U.S. Department of Energy's (DOE) Richland Operations Office to improve the waste management process to meet DOE's need for an efficient, cost-effective program for the management of dangerous, low-level and mixed-low-level waste. Additionally the program must meet all applicable regulatory requirements. The need for improvement was highlighted when a change in the Groundwater/Vadose Zone Integration Project's waste management practices resulted in a larger amount of waste being generated than the waste management organization had been set up to handle

Knowledge, attitudes and behaviour regarding waste management options in Romania: results from a school questionnaire

OpenAIRE

Karin KOLBE

2014-01-01

This study analyses knowledge, attitudes and behaviour in the area of different waste management approaches of pupils in Romania. Examining school students' knowledge about waste management options and finding out the reasons that prevent them from participating in environmentally sound disposal options is essential for teachers and legislators. For this purpose, questionnaires were designed and distributed in two schools in Romania. The analysis revealed that knowledge is highly developed in...

Results of research and development works 1981 of the Institute for Nuclear Waste Management Techniques

International Nuclear Information System (INIS)

1982-02-01

The emphasis of the work is on the development and characterization of waste products suitable for final storage, such as actinides and fission products from MAW/LAW, as well as on the development of methods for the treatment and solidification of radioactive wastes, e.g. wet chemical oxidation, vitrification and cementation. Investigations on the HAW-storage in salt are carried out in cooperation with the GSF. (HP) [de

Waste management at WAK

International Nuclear Information System (INIS)

Kuhn, K.D.; Willax, H.O.

1986-01-01

After a short description of the WAK plant and its reprocessing and intervention activities, types and sources of WAK wastes are described. Roughly half of the waste volume is generated during reprocessing, the other half during intervention periods. Most of the waste is transported to KfK for conditioning. Only waste from the head end cell is cementated on the spot. HLLW is stored in stainless steel tanks. Some results from analyzing this stuff are given. The corrosion behavior is acceptable for medium term storage. (orig.)

Results of Phase I groundwater quality assessment for single-shell tank waste management areas T and TX-TY at the Hanford Site

International Nuclear Information System (INIS)

Hodges, F.N.

1998-01-01

Pacific Northwest National Laboratory (PNNL) conducted a Phase I, Resource Conservation and Recovery Act of 1976 (RCRA) groundwater quality assessment for the Richland Field Office of the U.S. Department of Energy (DOE-RL) under the requirements of the Federal Facility Compliance Agreement. The purpose of the investigation was to determine if the Single-Shell Tank Waste Management Areas (WMAs) T and TX-TY have impacted groundwater quality. Waste Management Areas T and TX-TY, located in the northern part of the 200 West Area of the Hanford Site, contain the 241-T, 241-TX, and 241-TY tank farms and ancillary waste systems. These two units are regulated under RCRA interim-status regulations (under 40 CFR 265.93) and were placed in assessment groundwater monitoring because of elevated specific conductance in downgradient wells. Anomalous concentrations of technetium-99, chromium, nitrate, iodine-129, and cobalt-60 also were observed in some downgradient wells. Phase I assessment, allowed under 40 CFR 265, provides the owner-operator of a facility with the opportunity to show that the observed contamination has a source other than the regulated unit. For this Phase I assessment, PNNL evaluated available information on groundwater chemistry and past waste management practices in the vicinity of WMAs T and TX-TY. Background contaminant concentrations in the vicinity of WMAs T and TX-TY are the result of several overlapping contaminant plumes resulting from past-practice waste disposal operations. This background has been used as baseline for determining potential WMA impacts on groundwater

Chemodynamics of EDTA in a simulated mixed waste: the Hanford Site's complex concentrate waste

International Nuclear Information System (INIS)

Toste, A.P.; Ohnuki, Toshihiko

1999-01-01

Enormous stockpiles of mixed wastes at the USDOE's Hanford Site, the original US plutonium production facility, await permanent disposal. One mixed waste derived from reprocessing spent fuel was found to contain numerous nuclear related organics including chelating agents like EDTA and complexing agents, which have been used as decontamination agents, etc. Their presence in actual mixed wastes indicates that the organic content of nuclear wastes is dynamic and complicate waste management efforts. The subjects of this report is the chemo-degradation of EDTA degradation in a simulant Hanford's complex concentrate waste. The simulant was prepared by adding EDTA to an inorganic matrix, which was formulated based on past analyses of the actual waste. Aliquots of the EDTA simulant were withdrawn at different time points, derivatized via methylation and analyzed by gas chromatography and Gc/MS to monitor the disappearance of EDTA and the appearance of its' degradation products. This report also compares the results of EDTA's chemo-degradation to the g-radiolysis of EDTA in the simulant, the subject of a recently published article. Finally based on the results of these two studies, an assesment of the potential impact of EDTA degradation on the management of mixed wastes is offered. (J.P.N.)

Radioactive waste management