Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

Energy Technology Data Exchange (ETDEWEB)

Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-07-07

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

International Nuclear Information System (INIS)

Hansen, Francis D.; Kuhlman, Kristopher L.; Sobolik, Steven R.

2016-01-01

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

Radiolytic gas generation in salt cake technical task plan

International Nuclear Information System (INIS)

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

1993-01-01

High-level radioactive wastes are stored in large, steel tanks in the Savannah River Site Tank Farms. The liquid levels in these tanks are monitored to detect leakage of waste out of tanks or leakage of liquids into the tanks. Recent unexplained level fluctuations in high-level waste (HLW) tanks have caused High Level Waste Engineering (HLWE) to develop a program to better understand tank level behavior. Interim Waste Technology (IWT) has been requested by HLWE to obtain data which will lead to a better understanding of the radiolytic generations of gases in salt cake. The task described below will provide data from laboratory experiments with simulated wastes which can be used in tank level fluctuation modeling. The following experimental programs have been formulated to meet the task requirements of the customer: (A) determine whether radiolytically generated gas bubbles can be trapped in salt cake; (B) determine the composition of gases produced by radiolysis; (C) determine the yield of radiolysis gases as a function of radiation dose; (D) determine bubble distribution

Waste treatment using molten salt oxidation

International Nuclear Information System (INIS)

Navratil, J.D.; Stewart, A.E.

1996-01-01

MSO technology can be characterized as a submerged oxidation process; the basic concept is to introduce air and wastes into a bed of molten salt, oxidize the organic wastes in the molten salt, use the heat of oxidation to keep the salt molten and remove the salt for disposal or processing and recycling. The molten salt (usually sodium carbonate at 900-1000 C) provides four waste management functions: providing a heat transfer medium, catalyzing the oxidation reaction, preventing the formation of acid gases by forming stable salts, and efficiently capturing ash particles and radioactive materials by the combined effects of wetting, encapsulation and dissolution. The MSO process requires no wet scrubbing system for off-gas treatment. The process has been developed through bench-scale and pilot-scale testing, with successful destruction demonstration of a wide variety of hazardous and mixed (radioactive and hazardous wastes). (author). 24 refs, 2 tabs, 2 figs

Aspects on the gas generation and migration in repositories for high level waste in salt formations

International Nuclear Information System (INIS)

Ruebel, Andre; Buhmann, Dieter; Meleshyn, Artur; Moenig, Joerg; Spiessl, Sabine

2013-07-01

In a deep geological repository for high-level waste, gases may be produced during the post-closure phase by several processes. The generated gases can potentially affect safety relevant features and processes of the repository, like the barrier integrity, the transport of liquids and gases in the repository and the release of gaseous radionuclides from the repository into the biosphere. German long-term safety assessments for repositories for high-level waste in salt which were performed prior 2010 did not explicitly consider gas transport and the consequences from release of volatile radionuclides. Selected aspects of the generation and migration of gases in repositories for high-level waste in a salt formation are studied in this report from the viewpoint of the performance assessment. The knowledge on the availability of water in the repository, in particular the migration of salt rock internal fluids in the temperature field of the radioactive waste repository towards the emplacement drifts, was compiled and the amount of water was roughly estimated. Two other processes studied in this report are on the one hand the release of gaseous radionuclides from the repository and their potential impact in the biosphere and on the other hand the transport of gases along the drifts and shafts of the repository and their interaction with the fluid flow. The results presented show that there is some gas production expected to occur in the repository due to corrosion of container material from water disposed of with the backfill and inflowing from the host rock during the thermal phase. If not dedicated gas storage areas are foreseen in the repository concept, these gases might exceed the storage capacity for gases in the repository. Consequently, an outflow of gases from the repository could occur. If there are failed containers for spent fuel, radioactive gases might be released from the containers into the gas space of the backfill and subsequently transported together

Investigation of Various LiCl Waste Salt Purification Technologies

International Nuclear Information System (INIS)

Yung-Zun Cho; Hee-Chul Yang; Han-Soo Lee; In-Tae Kim

2008-01-01

Various purification research of LiCl waste molten salt generated from electroreduction process were tested. The purification of the LiCl waste salt very important in a various aspects, where the purification means separation of cesium and strontium form LiCl salt melts. In this study, for the separation of cesium and strontium from LiCl salt melts, precipitant agent addition techniques such as sulfate and carbonate addition method and, as a new attempt, zone freezing technique for concentration of cesium and strontium elements was investigated. As a results of this research, only strontium was carbonated by reaction with Li 2 CO 3 (cesium did not react with Li 2 CO 3 ). In case of sulfate addition method, both cesium and strontium were converted into their sulfate that is Cs 2 S 2 O 6 and SrSO 4 and maximum sulfate efficiency of cesium and strontium were about 72% and 95%, respectively. Cesium and strontium involved in LiCl molten salt could be concentrated in the molten salt by using zone freezing method. (authors)

Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report

International Nuclear Information System (INIS)

Miller, C.M.; Loomis, G.G.; Prewett, S.W.

1997-01-01

A proof-of-concept experimental study was performed to investigate the use of Orbit Technologies polysiloxane grouting material for encapsulation of U.S. Department of Energy mixed waste salts leading to a final waste form for disposal. Evaporator pond salt residues and other salt-like material contaminated with both radioactive isotopes and hazardous components are ubiquitous in the DOE complex and may exceed 250,000,000 kg of material. Current treatment involves mixing low waste percentages (less than 10% by mass salt) with cement or costly thermal treatment followed by cementation to the ash residue. The proposed technology involves simple mixing of the granular salt material (with relatively high waste loadings-greater than 50%) in a polysiloxane-based system that polymerizes to form a silicon-based polymer material. This study involved a mixing study to determine optimum waste loadings and compressive strengths of the resultant monoliths. Following the mixing study, durability testing was performed on promising waste forms. Leaching studies including the accelerated leach test and the toxicity characteristic leaching procedure were also performed on a high nitrate salt waste form. In addition to this testing, the waste form was examined by scanning electron microscope. Preliminary cost estimates for applying this technology to the DOE complex mixed waste salt problem is also given

Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

International Nuclear Information System (INIS)

Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

1997-01-01

Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE's mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies

Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

Energy Technology Data Exchange (ETDEWEB)

Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

1997-07-07

Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

Interim performance specifications for conceptual waste-package designs for geologic isolation in salt repositories

International Nuclear Information System (INIS)

1983-06-01

The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

Cementitious Stabilization of Mixed Wastes with High Salt Loadings

International Nuclear Information System (INIS)

Spence, R.D.; Burgess, M.W.; Fedorov, V.V.; Downing, D.J.

1999-01-01

Salt loadings approaching 50 wt % were tolerated in cementitious waste forms that still met leach and strength criteria, addressing a Technology Deficiency of low salt loadings previously identified by the Mixed Waste Focus Area. A statistical design quantified the effect of different stabilizing ingredients and salt loading on performance at lower loadings, allowing selection of the more effective ingredients for studying the higher salt loadings. In general, the final waste form needed to consist of 25 wt % of the dry stabilizing ingredients to meet the criteria used and 25 wt % water to form a workable paste, leaving 50 wt % for waste solids. The salt loading depends on the salt content of the waste solids but could be as high as 50 wt % if all the waste solids are salt

Computer simulation of an internally pressurized radioactive waste disposal room in a bedded salt formation

International Nuclear Information System (INIS)

Brown, W.T.; Weatherby, J.R.

1991-01-01

The Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico was created by the U.S. Department of Energy as an underground research and development facility to demonstrate the safe storage of transuranic waste generated from defense activities. This facility consists of storage rooms mined from a bedded salt formation at a depth of about 650 meters. Each room will accommodate about 6800 55-gallon drums filled with waste. After waste containers are emplaced, the storage rooms are to be backfilled with mined salt or other backfill materials. As time passes, reconsolidation of this backfill will reduce the hydraulic conductivity of the room. However, gases produced by decomposition and corrosion of waste and waste containers may cause a slow build-up of pressure which can retard consolidation of the waste and backfilled salt. The authors have developed a finite-element model of an idealized disposal room which is assumed to be perfectly sealed. The assumption that no gas escapes from the disposal room is a highly idealized and extreme condition which does not account for leakage paths, such as interbeds, that exist in the surrounding salt formation. This model has been used in a parametric study to determine how reconsolidation is influenced by various assumed gas generation rates and total amounts of gas generated. Results show that reductions in the gas generation, relative to the baseline case, can increase the degree of consolidation and reduce the peak gas pressure in disposal rooms. Even higher degrees of reconsolidation can be achieved by reducing both amounts and rates of gas generation. 8 refs., 4 figs., 1 tab

Molt salts reactors capacity for wastes incineration and energy production

International Nuclear Information System (INIS)

David, S.; Nuttin, A.

2005-01-01

The molten salt reactors present many advantages in the framework of the IV generation systems development for the energy production and/or the wastes incineration. After a recall of the main studies realized on the molten salt reactors, this document presents the new concepts and the identified research axis: the MSRE project and experience, the incinerators concepts, the thorium cycle. (A.L.B.)

Deep geologic disposal of mixed waste in bedded salt: The Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Rempe, N.T.

1993-01-01

Mixed waste (i.e., waste that contains both chemically hazardous and radioactive components) poses a moral, political, and technical challenge to present and future generations. But an international consensus is emerging that harmful byproducts and residues can be permanently isolated from the biosphere in a safe and environmentally responsible manner by deep geologic disposal. To investigate and demonstrate such disposal for transuranic mixed waste, derived from defense-related activities, the US Department of Energy has prepared the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. This research and development facility was excavated approximately at the center of a 600 m thick sequence of salt (halite) beds, 655 m below the surface. Proof of the long-term tectonic and hydrological stability of the region is supplied by the fact that these salt beds have remained essentially undisturbed since they were deposited during the Late Permian age, approximately 225 million years ago. Plutonium-239, the main radioactive component of transuranic mixed waste, has a half-life of 24,500 years. Even ten half-lives of this isotope - amounting to about a quarter million years, the time during which its activity will decline to background level represent only 0.11 percent of the history of the repository medium. Therefore, deep geologic disposal of transuranic mixed waste in Permian bedded salt appears eminently feasible

Alternatives for definse waste-salt disposal

International Nuclear Information System (INIS)

Benjamin, R.W.; McDonell, W.R.

1983-01-01

Alternatives for disposal of decontaminated high-level waste salt at Savannah River were reviewed to estimate costs and potential environmental impact for several processes. In this review, the reference process utilizing intermediate-depth burial of salt-concrete (saltcrete) monoliths was compared with alternatives including land application of the decontaminated salt as fertilizer for SRP pine stands, ocean disposal with and without containment, and terminal storage as saltcake in existing SRP waste tanks. Discounted total costs for the reference process and its modifications were in the same range as those for most of the alternative processes; uncontained ocean disposal with truck transport to Savannah River barges and storage as saltcake in SRP tanks had lower costs, but presented other difficulties. Environmental impacts could generally be maintained within acceptable limits for all processes except retention of saltcake in waste tanks, which could result in chemical contamination of surrounding areas on tank collapse. Land application would require additional salt decontamination to meet radioactive waste disposal standards, and ocean disposal without containment is not permitted in existing US practice. The reference process was judged to be the only salt disposal option studied which would meet all current requirements at an acceptable cost

Organic waste processing using molten salt oxidation

Energy Technology Data Exchange (ETDEWEB)

Adamson, M. G., LLNL

1998-03-01

Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

X-ray diffraction of slag-based sodium salt waste forms

Energy Technology Data Exchange (ETDEWEB)

Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-09-30

The attached report documents sample preparation and x-ray diffraction results for a series of cement and blended cement matrices prepared with either water or a 4.4 M Na salt solution. The objective of the study was to provide initial phase characterization for the Cementitious Barriers Partnership reference case cementitious salt waste form. This information can be used to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) potentially to design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available.

Test procedures for polyester immobilized salt-containing surrogate mixed wastes

International Nuclear Information System (INIS)

Biyani, R.K.; Hendrickson, D.W.

1997-01-01

These test procedures are written to meet the procedural needs of the Test Plan for immobilization of salt containing surrogate mixed waste using polymer resins, HNF-SD-RE-TP-026 and to ensure adequacy of conduct and collection of samples and data. This testing will demonstrate the use of four different polyester vinyl ester resins in the solidification of surrogate liquid and dry wastes, similar to some mixed wastes generated by DOE operations

Some geotechnical problems related to underground waste disposal in salt formations

International Nuclear Information System (INIS)

Berest, P.

1993-01-01

Nuclear waste disposal in deep salt formations is an option considered by several countries. Rock salt is a very impervious medium, but can be easily leached; selection of an appropriate disposal formation must account for natural protections of the formation as regards water movements. It must be checked that such initially favourable characteristics will not be affected by the existence of shafts and galleries, or by the important heat output generated by vitrified wastes. The discussion is uneasy, for a comprehensive rheological model for rock salt is difficult to set and to be extrapolated to large time scales; some methodological problems are raised by use of numerical computations. (author). 22 refs., 2 figs

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.

A Dual-Continuum Model for Brine Migration in Salt Associated with Heat-Generating Nuclear Waste: Fully Coupled Thermal-Hydro-Mechanical Analysis

Science.gov (United States)

Hu, M.; Rutqvist, J.

2017-12-01

The disposal of heat-generating nuclear waste in salt host rock establishes a thermal gradient around the waste package that may cause brine inclusions in the salt grains to migrate toward the waste package. In this study, a dual-continuum model is developed to analyze such a phenomenon. This model is based on the Finite Volume Method (FVM), and it is fully thermal-hydro-mechanical (THM) coupled. For fluid flow, the dual-continuum model considers flow in the interconnected pore space and also in the salt grains. The mass balance of salt and water in these two continua is separately established, and their coupling is represented by flux associated with brine migration. Together with energy balance, such a system produces a coupled TH model with strongly nonlinear features. For mechanical analysis, a new formulation is developed based on the Voronoi tessellated mesh. By relating each cell to several connected triangles, first-order approximation is constructed. The coupling between thermal and mechanical fields is only considered in terms of thermal expansion. And the coupling between the hydraulic and mechanical fields in terms of pore-volume effects is consistent with Biot's theory. Therefore, a fully coupled THM model is developed. Several demonstration examples are provided to verify the model. Last the new model is applied to analyze coupled THM behavior and the results are compared with experimental data.

Disposal of Savannah River Plant waste salt

International Nuclear Information System (INIS)

Dukes, M.D.

1982-01-01

Approximately 26-million gallons of soluble low-level waste salts will be produced during solidification of 6-million gallons of high-level defense waste in the proposed Defense Waste Processing Facility (DWPF) at the Savannah River Plant (SRP). Soluble wastes (primarily NaNO 3 , NaNO 2 , and NaOH) stored in the waste tanks will be decontaminated by ion exchange and solidified in concrete. The resulting salt-concrete mixture, saltcrete, will be placed in a landfill on the plantsite such that all applicable federal and state disposal criteria are met. Proposed NRC guidelines for the disposal of waste with the radionuclide content of SRP salt would permit shallow land burial. Federal and state rules require that potentially hazardous chemical wastes (mainly nitrate-nitrate salts in the saltcrete) be contained to the degree necessary to meet drinking water standards in the ground water beneath the landfill boundary. This paper describes the proposed saltcrete landfill and tests under way to ensure that the landfill will meet these criteria. The work includes laboratory and field tests of the saltcrete itself, a field test of a one-tenth linear scale model of the entire landfill system, and a numerical model of the system

Development and characterization of new high-level waste form containing LiCl KCl eutectic salts for achieving waste minimization from pyroprocessing

International Nuclear Information System (INIS)

Cho, Yong Zun; Kim, In Tae; Park, Hwan Seo; Ahn, Byeung Gil; Eun, Hee Chul; Son, Seock Mo; Ah, Su Na

2011-12-01

The purpose of this project is to develop new high level waste (HLW) forms and fabrication processes to dispose of active metal fission products that are removed from electrorefiner salts in the pyroprocessing based fuel cycle. The current technology for disposing of active metal fission products in pyroprocessing involves non selectively discarding of fission product loaded salt in a glass-bonded sodalite ceramic waste form. Selective removal of fission products from the molten salt would greatly minimize the amount of HLW generated and methods were developed to achieve selective separation of fission products during a previous I NERI research project (I NERI 2006 002 K). This I NERI project proceeds from the previous project with the development of suitable waste forms to immobilize the separated fission products. The Korea Atomic Energy Research Institute (KAERI) has focused primarily on developing these waste forms using surrogate waste materials, while the Idaho National Laboratory (INL) has demonstrated fabrication of these waste forms using radioactive electrorefiner salts in hot cell facilities available at INL. Testing and characterization of these radioactive materials was also performed to determine the physical, chemical, and durability properties of the waste forms

Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

Science.gov (United States)

Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

2018-04-01

In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

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.

Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-06-01

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

Solid waste disposal into salt mines

International Nuclear Information System (INIS)

Repke, W.

1981-01-01

The subject is discussed as follows: general introduction to disposal of radioactive waste; handling of solid nuclear waste; technology of final disposal, with specific reference to salt domes; conditioning of radioactive waste; safety barriers for radioactive waste; practice of final disposal in other countries. (U.K.)

Concepts and Technologies for Radioactive Waste Disposal in Rock Salt

Directory of Open Access Journals (Sweden)

Wernt Brewitz

2007-01-01

Full Text Available In Germany, rock salt was selected to host a repository for radioactive waste because of its excellent mechanical properties. During 12 years of practical disposal operation in the Asse mine and 25 years of disposal in the disused former salt mine Morsleben, it was demonstrated that low-level wastes (LLW and intermediate-level wastes (ILW can be safely handled and economically disposed of in salt repositories without a great technical effort. LLW drums were stacked in old mining chambers by loading vehicles or emplaced by means of the dumping technique. Generally, the remaining voids were backfilled by crushed salt or brown coal filter ash. ILW were lowered into inaccessible chambers through a borehole from a loading station above using a remote control.Additionally, an in-situ solidification of liquid LLW was applied in the Morsleben mine. Concepts and techniques for the disposal of heat generating high-level waste (HLW are advanced as well. The feasibility of both borehole and drift disposal concepts have been proved by about 30 years of testing in the Asse mine. Since 1980s, several full-scale in-situ tests were conducted for simulating the borehole emplacement of vitrified HLW canisters and the drift emplacement of spent fuel in Pollux casks. Since 1979, the Gorleben salt dome has been investigated to prove its suitability to host the national final repository for all types of radioactive waste. The “Concept Repository Gorleben” disposal concepts and techniques for LLW and ILW are widely based on the successful test operations performed at Asse. Full-scale experiments including the development and testing of adequate transport and emplacement systems for HLW, however, are still pending. General discussions on the retrievability and the reversibility are going on.

Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

International Nuclear Information System (INIS)

Hsu, P.C.

1997-01-01

Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment

Conceptual waste package interim product specifications and data requirements for disposal of glass commercial high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-10-01

The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses and regulatory requirements become available. 13 references, 1 figure

Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

Energy Technology Data Exchange (ETDEWEB)

Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Trone, Janis R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Lawrence C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-12-01

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

Molten salt treatment to minimize and optimize waste

International Nuclear Information System (INIS)

Gat, U.; Crosley, S.M.; Gay, R.L.

1993-01-01

A combination molten salt oxidizer (MSO) and molten salt reactor (MSR) is described for treatment of waste. The MSO is proposed for contained oxidization of organic hazardous waste, for reduction of mass and volume of dilute waste by evaporation of the water. The NTSO residue is to be treated to optimize the waste in terms of its composition, chemical form, mixture, concentration, encapsulation, shape, size, and configuration. Accumulations and storage are minimized, shipments are sized for low risk. Actinides, fissile material, and long-lived isotopes are separated and completely burned or transmuted in an MSR. The MSR requires no fuel element fabrication, accepts the materials as salts in arbitrarily small quantities enhancing safety, security, and overall acceptability

Mixing Modeling Analysis For SRS Salt Waste Disposition

International Nuclear Information System (INIS)

Lee, S.

2011-01-01

Nuclear waste at Savannah River Site (SRS) waste tanks consists of three different types of waste forms. They are the lighter salt solutions referred to as supernate, the precipitated salts as salt cake, and heavier fine solids as sludge. The sludge is settled on the tank floor. About half of the residual waste radioactivity is contained in the sludge, which is only about 8 percentage of the total waste volume. Mixing study to be evaluated here for the Salt Disposition Integration (SDI) project focuses on supernate preparations in waste tanks prior to transfer to the Salt Waste Processing Facility (SWPF) feed tank. The methods to mix and blend the contents of the SRS blend tanks were evalutaed to ensure that the contents are properly blended before they are transferred from the blend tank such as Tank 50H to the SWPF feed tank. The work consists of two principal objectives to investigate two different pumps. One objective is to identify a suitable pumping arrangement that will adequately blend/mix two miscible liquids to obtain a uniform composition in the tank with a minimum level of sludge solid particulate in suspension. The other is to estimate the elevation in the tank at which the transfer pump inlet should be located where the solid concentration of the entrained fluid remains below the acceptance criterion (0.09 wt% or 1200 mg/liter) during transfer operation to the SWPF. Tank 50H is a Waste Tank that will be used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work described here consists of two modeling areas. They are the mixing modeling analysis during miscible liquid blending operation, and the flow pattern analysis during transfer operation of the blended liquid. The modeling results will provide quantitative design and operation information during the mixing/blending process and the transfer operation of the blended

Waste salt recovery, recycle, and destruction

International Nuclear Information System (INIS)

Hickman, R.G.

1992-12-01

Starting in 1943 and continuing into the 1970s, radioactive wastes resulting from plutonium processing at Hanford were stored underground in 149 single shell tanks. Of these tanks, 66 are known or believedto be leaking, and over a period are believed to have leaked about 750,000 gal into the surrounding soil. The bulk of the aqueous solution has been removed and transferred to double shell tanks, none of which are leaking. The waste consists of 37 million gallons of salt cake and sludge. Most of the salt cake is sodium nitrate and other sodium salts. A substantial fraction of the sludge is sodium nitrate. Small amounts of the radionuclides are present in the sludge as oxides or hydroxides. In addition, some of the tanks contain organic compounds and ferrocyanide complexes, many of which have undergone radiolytic induced chemical changes during the years of storage. As part of the Hanford site remediation effort, the tank wastes must be removed, treated, and the residuals must be immobilized and disposed of in an environmentally acceptable manner. Removal methods of the waste from the tanks fall generally into three approaches: dry removal, slurry removal, and solution removed. The latter two methods are likely to result in some additional leakage to the surrounding soil, but that may be acceptable if the tank can be emptied and remediated before the leaked material permeates deeply into the soil. This effort includes three parts: salt splitting, acid separation, and destruction, with initial emphasis on salt splitting

Molten salt hazardous waste disposal process utilizing gas/liquid contact for salt recovery

International Nuclear Information System (INIS)

Grantham, L.F.; McKenzie, D.E.

1984-01-01

The products of a molten salt combustion of hazardous wastes are converted into a cooled gas, which can be filtered to remove hazardous particulate material, and a dry flowable mixture of salts, which can be recycled for use in the molten salt combustion, by means of gas/liquid contact between the gaseous products of combustion of the hazardous waste and a solution produced by quenching the spent melt from such molten salt combustion. The process results in maximizing the proportion of useful materials recovered from the molten salt combustion and minimizing the volume of material which must be discarded. In a preferred embodiment a spray dryer treatment is used to achieve the desired gas/liquid contact

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Geologic disposal of nuclear wastes: salt's lead is challenged

International Nuclear Information System (INIS)

Kerr, R.A.

1979-01-01

The types of radioactive waste disposal sites available are outlined. The use of salt deposits and their advantages are discussed. The reasons for the selection of the present site for the Waste Isolation Pilot Plant are presented. The possibilities of using salt domes along the Gulf Coast and not-salt rocks as nuclear waste repositories are also discussed. The sea bed characteristics are described and advantages of this type of site selection are presented

Radiolytic gas generation in Salt Cake Quality Assurance Plan

International Nuclear Information System (INIS)

Walker, D.D.

1993-01-01

High-level radioactive wastes are stored in large, steel tanks in the Savannah River Site's Tank Farms. The liquid levels in these tanks are monitored to detect leakage of waste out of tanks or leakage of liquids into the tanks. Recent unexplained level fluctuations in high-level waste (HLW) tanks have caused High Level Waste Engineering (HLWE) to develop a program to better understand tank level behavioral Interim Waste Technology (IWT) has been requested by HLWE to obtain data which will lead to a better understanding of the radiolytic generation of gases in salt cake. A task plan has been written in response to this request. This document details the controls necessary to ensure the quality of the results of the activities described in the task plan

Expected environment for waste packages in a salt repository

International Nuclear Information System (INIS)

Pederson, L.R.; Clark, D.E.; Hodges, F.N.; McVay, G.L.; Rai, D.

1983-01-01

This paper discusses results of recent efforts to define the very near-field (within approximately 2 m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and standard brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported. 38 references, 3 figures, 2 tables

Molten salt processing of mixed wastes with offgas condensation

International Nuclear Information System (INIS)

Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R.; Gay, R.L.; Stewart, A.; Yosim, S.

1991-01-01

We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000 degrees C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700 degrees C. 15 refs., 5 figs., 1 tab

Radioactive waste and special waste disposal in salt domes - phoney waste management solutions

International Nuclear Information System (INIS)

Grimmel, E.

1990-01-01

The paper tries to make aware of the fact that an indefinite safe disposal of anthropogeneous wastes in underground repositories is impossible. Suspicion is raised that the Gorleben-Rambow salt dome has never been studied for its suitability as a repository, but that it was simply taken for granted. Safety analyses are meant only to conceal uncertainty. It is demanded to immediately opt out of the ultimate disposal technique for radioactive and special wastes in salt caverns. (DG) [de

Radioactive waste generated from JAERI partitioning-transmutation cycle system

Energy Technology Data Exchange (ETDEWEB)

Shinichi, Nakayama; Yasuji, Morita; Kenji, Nishihara [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2001-07-01

Production of lower-level radioactive wastes, as well as the reduction in radioactivity of HLW, is an important performance indicator in assessing the viability of a partitioning-transmutation system. We have begun to identify the chemical compositions and to quantify the amounts of radioactive wastes that may be generated by JAERI processes. Long-lived radionuclides such as {sup 14}C and {sup 59}Ni and spallation products of Pb-Bi coolants are added to the existing inventory of these nuclides that are generated in the current fuel cycle. Spent salts of KCl-LiCl, which is not generated from the current fuel cycle, will be introduced as a waste. (author)

Radioactive Wastes Generated From JAERI Partitioning-Transmutation Fuel Cycle

International Nuclear Information System (INIS)

Nakayama, Shinichi; Morita, Yasuji; Nishihara, Kenji

2003-01-01

Production of lower-level radioactive wastes, as well as the reduction in radioactivity of HLW, is an important performance indicator in assessing the viability of a partitioning-transmutation system. We have begun to identify the chemical compositions and to quantify the amounts of radioactive wastes that may be generated by JAERI's processes. Long-lived radionuclides such as 14 C and 59 Ni and spallation products of Pb-Bi coolants are added to the existing inventory of these nuclides that are generated in the current fuel cycle. Spent salts of KCl-LiCl, which is not generated from the current fuel cycle, will be introduced as a waste. (authors)

Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

International Nuclear Information System (INIS)

Dmitriev, S. A.; Adamovich, D. V.; Demkin, V. I.; Timofeev, E. M.

2003-01-01

The main fields of introduction and application of membrane methods for preliminary treatment and processing salt liquid radioactive waste (SLRW) can be nuclear power stations (NPP) and enterprises on atomic submarines (AS) utilization. Unlike the earlier developed technology for the liquid salt bearing radioactive waste decontamination and concentrating this report presents the new enhanced membrane technology for the liquid salt bearing radioactive waste processing based on the state-of-the-art membrane unit design, namely, the filtering units equipped with the metal-ceramic membranes of ''TruMem'' brand, as well as the electrodialysis and electroosmosis concentrators. Application of the above mentioned units in conjunction with the pulse pole changer will allow the marked increase of the radioactive waste concentrating factor and the significant reduction of the waste volume intended for conversion into monolith and disposal. Besides, the application of the electrodialysis units loaded with an ion exchange material at the end polishing stage of the radioactive waste decontamination process will allow the reagent-free radioactive waste treatment that meets the standards set for the release of the decontaminated liquid radioactive waste effluents into the natural reservoirs of fish-farming value

Immobilization of IFR salt wastes in mortar

International Nuclear Information System (INIS)

Fischer, D.F.; Johnson, T.R.

1988-01-01

Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered

Salt removal from tanks containing high-level radioactive waste

International Nuclear Information System (INIS)

Kiser, D.L.

1981-01-01

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

Recovery of soluble chloride salts from the wastewater generated during the washing process of municipal solid wastes incineration fly ash.

Science.gov (United States)

Tang, Hailong; Erzat, Aris; Liu, Yangsheng

2014-01-01

Water washing is widely used as the pretreatment method to treat municipal solid waste incineration fly ash, which facilitates the further solidification/stabilization treatment or resource recovery of the fly ash. The wastewater generated during the washing process is a kind of hydrosaline solution, usually containing high concentrations of alkali chlorides and sulphates, which cause serious pollution to environment. However, these salts can be recycled as resources instead of discharge. This paper explored an effective and practical recovery method to separate sodium chloride, potassium chloride, and calcium chloride salts individually from the hydrosaline water. In laboratory experiments, a simulating hydrosaline solution was prepared according to composition of the waste washing water. First, in the three-step evaporation-crystallization process, pure sodium chloride and solid mixture of sodium and potassium chlorides were obtained separately, and the remaining solution contained potassium and calcium chlorides (solution A). And then, the solid mixture was fully dissolved into water (solution B obtained). Finally, ethanol was added into solutions A and B to change the solubility of sodium, potassium, and calcium chlorides within the mixed solvent of water and ethanol. During the ethanol-adding precipitation process, each salt was separated individually, and the purity of the raw production in laboratory experiments reached about 90%. The ethanol can be recycled by distillation and reused as the solvent. Therefore, this technology may bring both environmental and economic benefits.

Saltstone: cement-based waste form for disposal of Savannah River Plant low-level radioactive salt waste

International Nuclear Information System (INIS)

Langton, C.A.

1984-01-01

Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO 3 , NaOH, Na 2 SO 4 , and NaNO 2 . After decontamination, the salt solution is classified as low-level waste. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses. The disposal system for the DWPF salt waste includes reconstitution of the crystallized salt as a solution containing 32 wt % solids. This solution will be decontaminated to remove 137 Cs and 90 Sr and then stabilized in a cement-based waste form. Laboratory and field tests indicate that this stabilization process greatly reduces the mobility of all of the waste constitutents in the surface and near-surface environment. Engineered trenches for subsurface burial of the saltstone have been designed to ensure compatibility between the waste form and the environment. The total disposal sytem, saltstone-trench-surrounding soil, has been designed to contain radionuclides, Cr, and Hg by both physical encapsulation and chemical fixation mechanisms. Physical encapsulation of the salts is the mechanism employed for controlling N and OH releases. In this way, final disposal of the SRP low-level waste can be achieved and the quality of the groundwater at the perimeter of the disposal site meets EPA drinking water standards

Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

International Nuclear Information System (INIS)

1999-01-01

Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

Criticality considerations for salt-cake disolution in DOE waste tanks

International Nuclear Information System (INIS)

Trumble, E.F.; Niemer, K.A.

1995-01-01

A large amount of high-level waste is being stored in the form of salt cake at the Savannah River site (SRS) in large (1.3 x 106 gal) underground tanks awaiting startup of the Defense Waste Processing Facility (DWPF). This salt cake will be dissolved with water, and the solution will be fed to DWPF for immobilization in borosilicate glass. Some of the waste that was transferred to the tanks contained enriched uranium and plutonium from chemical reprocessing streams. As water is added to these tanks to dissolve the salt cake, the insoluble portion of this fissile material will be left behind in the tank as the salt solution is pumped out. Because the salt acts as a diluent to the fissile material, the process of repeated water addition, salt dissolution, and salt solution removal will act as a concentrating mechanism for the undissolved fissile material that will remain in the tank. It is estimated that tank 41 H at SRS contains 20 to 120 kg of enriched uranium, varying from 10 to 70% 235 U, distributed nonuniformly throughout the tank. This paper discusses the criticality concerns associated with the dissolution of salt cake in this tank. These concerns are also applicable to other salt cake waste tanks that contain significant quantities of enriched uranium and/or plutonium

Release rates from waste packages in a salt repository

International Nuclear Information System (INIS)

Chambre, P.L.; Hwang, Y.; Lee, W.W.L.; Pigford, T.H.

1987-06-01

In this report we present estimates of radionuclide release rates from waste packages into salt. This conservative and bounding analysis shows that release rates from waste packages in salt are well below the US Nuclear Regulatory Commission's performance objectives for the engineered barrier system. 2 refs., 2 figs

Hydrothermal preparation of zeolite Li-A and ion exchange properties of Cs and Sr in salt waste

International Nuclear Information System (INIS)

Lee, S. H.; Kim, J. G.; Lee, J. H.; Kim, J. H.

2005-01-01

An advanced spent fuel management process that were based on Li reduction of the oxide spent fuel to a metallic form will generate a LiCl waste. Zeolite A has been reported as a promising immobilization medium for waste salt with CsCl and SrCl 2 . However, Sodium is accumulated as an ionic form (Na + -ion) in molten salt during ion exchange step between Na + -ion in zeolite A and Li + -ion in the molten salt. Therefore, zeolite Na-A need to be replaced by the Li-type zeolite for recycling the salt waste by removing the Cs and Sr ions. In this study, the hydrothermal preparation of zeolite Li-A was performed in 350ml pressure vessel by P. Norby method. The preparation characteristics of zeolite Li-A was investigated. And the ion exchange properties of Cs and Sr in molten LiCl salt were investigated under the condition of 923K using zeolite 4A and prepared zeolite Li-A

Laboratory simulation of salt dissolution during waste removal

International Nuclear Information System (INIS)

Wiersma, B.J.; Parish, W.R.

1997-01-01

Laboratory experiments were performed to support the field demonstration of improved techniques for salt dissolution in waste tanks at the Savannah River Site. The tests were designed to investigate three density driven techniques for salt dissolution: (1) Drain-Add-Sit-Remove, (2) Modified Density Gradient, and (3) Continuous Salt Mining. Salt dissolution was observed to be a very rapid process as salt solutions with densities between 1.38-1.4 were frequently removed. Slower addition and removal rates and locating the outlet line at deeper levels below the top of the saltcake provided the best contact between the dissolution water and the saltcake. It was observed that dissolution with 1 M sodium hydroxide solution resulted in salt solutions that were within the current inhibitor requirements for the prevention of stress corrosion cracking. This result was independent of the density driven technique. However, if inhibited water (0.01 M sodium hydroxide and 0.011 M sodium nitrite) was utilized, the salt solutions were frequently outside the inhibitor requirements. Corrosion testing at conditions similar to the environments expected during waste removal was recommended

Alternative methods of salt disposal at the seven salt sites for a nuclear waste repository

International Nuclear Information System (INIS)

1987-02-01

This study discusses the various alternative salt management techniques for the disposal of excess mined salt at seven potentially acceptable nuclear waste repository sites: Deaf Smith and Swisher Counties, Texas; Richton and Cypress Creek Domes, Mississippi; Vacherie Dome, Louisiana; and Davis and Lavender Canyons, Utah. Because the repository development involves the underground excavation of corridors and waste emplacement rooms, in either bedded or domed salt formations, excess salt will be mined and must be disposed of offsite. The salt disposal alternatives examined for all the sites include commercial use, ocean disposal, deep well injection, landfill disposal, and underground mine disposal. These alternatives (and other site-specific disposal methods) are reviewed, using estimated amounts of excavated, backfilled, and excess salt. Methods of transporting the excess salt are discussed, along with possible impacts of each disposal method and potential regulatory requirements. A preferred method of disposal is recommended for each potentially acceptable repository site. 14 refs., 5 tabs

Waste management analysis for the nuclear fuel cycle. I. Actinide recovery from aqueous salt wastes

International Nuclear Information System (INIS)

Martella, L.L.; Navratil, J.D.

1979-01-01

A preliminary feasibility study of solvent extraction methods has been completed for removing actinides from selected salt wastes likely to be produced during reactor fuel fabrication and reprocessing. The use of a two-step solvent extraction system, tributyl phosphate (TBP) followed by a bidentate organophosphorus extractant (DHDECMP), appears most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and >99.99% of the uranium. The second step, using DHDECMP, would remove >99.91% of the americium, the remaining plutonium (>99.98%), and other actinides from the acidified salt waste

Defense waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

Langton, C.A.; Dukes, M.D.

1984-01-01

A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. The disposal process includes emplacing the saltstone in engineered trenches above the water table but below grade at SRP. Design of the waste form and disposal system limits the concentration of salts and radionuclides in the groundwater so that EPA drinking water standards will not be exceeded at the perimeter of the disposal site. 10 references, 4 figures, 3 tables

Thermal performances of molten salt steam generator

International Nuclear Information System (INIS)

Yuan, Yibo; He, Canming; Lu, Jianfeng; Ding, Jing

2016-01-01

Highlights: • Thermal performances of molten salt steam generator were experimentally studied. • Overall heat transfer coefficient reached maximum with optimal molten salt flow rate. • Energy efficiency first rose and then decreased with salt flow rate and temperature. • Optimal molten salt flow rate and temperature existed for good thermal performance. • High inlet water temperature benefited steam generating rate and energy efficiency. - Abstract: Molten salt steam generator is the key technology for thermal energy conversion from high temperature molten salt to steam, and it is used in solar thermal power station and molten salt reactor. A shell and tube type molten salt steam generator was set up, and its thermal performance and heat transfer mechanism were studied. As a coupling heat transfer process, molten salt steam generation is mainly affected by molten salt convective heat transfer and boiling heat transfer, while its energy efficiency is also affected by the heat loss. As molten salt temperature increased, the energy efficiency first rose with the increase of heat flow absorbed by water/steam, and then slightly decreased for large heat loss as the absorbed heat flow still rising. At very high molten salt temperature, the absorbed heat flow decreased as boiling heat transfer coefficient dropping, and then the energy efficiency quickly dropped. As the inlet water temperature increased, the boiling region in the steam generator remarkably expanded, and then the steam generation rate and energy efficiency both rose with the overall heat transfer coefficient increasing. As the molten salt flow rate increased, the wall temperature rose and the boiling heat transfer coefficient first increased and then decreased according to the boiling curve, so the overall heat transfer coefficient first increased and then decreased, and then the steam generation rate and energy efficiency of steam generator both had maxima.

Expedited demonstration of molten salt mixed waste treatment technology. Final report

International Nuclear Information System (INIS)

1995-01-01

This final report discusses the molten salt mixed waste project in terms of the various subtasks established. Subtask 1: Carbon monoxide emissions; Establish a salt recycle schedule and/or a strategy for off-gas control for MWMF that keeps carbon monoxide emission below 100 ppm on an hourly averaged basis. Subtask 2: Salt melt viscosity; Experiments are conducted to determine salt viscosity as a function of ash composition, ash concentration, temperature, and time. Subtask 3: Determine that the amount of sodium carbonate entrained in the off-gas is minimal, and that any deposited salt can easily be removed form the piping using a soot blower or other means. Subtask 4: The provision of at least one final waste form that meets the waste acceptance criteria of a landfill that will take the waste. This report discusses the progress made in each of these areas

Salt Repository Project Waste Package Program Plan: Draft

International Nuclear Information System (INIS)

Carr, J.A.; Cunnane, J.C.

1986-01-01

Under the direction of the Office of Civilian Radioactive Waste Management (OCRWM) created within the DOE by direction of the Nuclear Waste Policy Act of 1982 (NWPA), the mission of the Salt Repository Project (SRP) is to provide for the development of a candidate salt repository for disposal of high-level radioactive waste (HLW) and spent reactor fuel in a manner that fully protects the health and safety of the public and the quality of the environment. In consideration of the program needs and requirements discussed above, the SRP has decided to develop and issue this SRP Waste Package Program Plan. This document is intended to outline how the SRP plans to develop the waste package design and to show, with reasonable assurance, that the developed design will satisfy applicable requirements/performance objectives. 44 refs., 16 figs., 16 tabs

Engineering study of the potential uses of salts from selective crystallization of Hanford tank wastes

International Nuclear Information System (INIS)

Hendrickson, D.W.

1996-01-01

The Clean Salt Process (CSP) is the fractional crystallization of nitrate salts from tank waste stored on the Hanford Site. This study reviews disposition options for a CSP product made from Hanford Site tank waste. These options range from public release to onsite low-level waste disposal to no action. Process, production, safety, environment, cost, schedule, and the amount of CSP material which may be used are factors considered in each option. The preferred alternative is offsite release of clean salt. Savings all be generated by excluding the material from low-level waste stabilization. Income would be received from sales of salt products. Savings and income from this alternative amount to $1,027 million, excluding the cost of CSP operations. Unless public sale of CSP products is approved, the material should be calcined. The carbonate form of the CSP could then be used as ballast in tank closure and stabilization efforts. Not including the cost of CSP operations, savings of $632 million would be realized. These savings would result from excluding the material from low-level waste stabilization and reducing purchases of chemicals for caustic recycle and stabilization and closure. Dose considerations for either alternative are favorable. No other cost-effective alternatives that were considered had the capacity to handle significant quantities of the CSP products. If CSP occurs, full-scale tank-waste stabilization could be done without building additional treatment facilities after Phase 1 (DOE 1996). Savings in capital and operating cost from this reduction in waste stabilization would be in addition to the other gains described

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.)

Salt splitting of sodium-dominated radioactive waste using ceramic membranes

International Nuclear Information System (INIS)

Hollenberg, G.W.; Carlson, C.D.; Virkar, A.; Joshi, A.

1994-08-01

The potential for salt splitting of sodium dominated radioactive wastes by use of a ceramic membrane is reviewed. The technical basis for considering this processing technology is derived from the technology developed for battery and chlor-alkali chemical industry. Specific comparisons are made with the commercial organic membranes which are the standard in nonradioactive salt splitting. Two features of ceramic membranes are expected to be especially attractive: high tolerance to gamma irradiation and high selectivity between sodium and other ions. The objective of the salt splitting process is to separate nonradioactive sodium from contaminated sodium salts prior to other pretreatment processes in order to: (1) concentrate the waste in order to reduce the volume of subsequent additives and capacity of equipment, (2) decrease the pH of the waste in preparation for further processing, and (3) provide sodium with very low radioactivity levels for caustic washing of sludge or low level and mixed waste vitrification

Molten salt destruction process for mixed wastes

International Nuclear Information System (INIS)

Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

1993-04-01

We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

Containment of solidified liquid hazardous waste in domal salt

International Nuclear Information System (INIS)

Domenico, P.A.; Lerman, A.

1992-01-01

In recent years, the solidification of hazardous liquid waste has become a viable option in waste management. The solidification process results in an increased volume but more stable waste form that must be disposed of or stored in a dry environment. An environment of choice in south central Texas is domal salt. The salt dome currently under investigation has a water content of 0.002 percent by weight and a permeability less than one nanodarcy. A question that must be addressed is whether a salt dome has a particular set of attributes that will prevent the release of contaminants to the environment. From a regulatory perspective, a ''no migration'' petition must be approved by the U.S.E.P.A. for the containment facility. By ''no migration'' it is implied that the waste must be contained for 10,000 years. A demonstration that this condition will be met will require model calculations and such models must be based on the physical and chemical characteristics of the waste form and the geologic environment. In particular, the models must address the rate of brine infiltration into the caverns, providing information on how fast an immobile solid waste form could convert to a more mobile liquid state. Additionally, the potential for migration by both diffusion and advection is of concern. Lastly, given a partially saturated cavern, the question of how far gaseous waste will be transported over the 10,000 year containment period must also be addressed. Results indicate that the containment capabilities of domal salt are exceptional. A nominal volume of brine will seep into the cavern and most voids between the injected solidified waste pellets will remain unsaturated. Very small quantities of hazardous constituents will be leached from the waste pellets

Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.5

International Nuclear Information System (INIS)

1981-01-01

The present report deals with safety evaluation as part of the investigations regarding a repository for high-level waste in a salt dome. It is volume 5 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. Two characteristics of the waste are of special importance for the safety evaluation: the encasing of the waste in steel casks with 15 cm thick walls affording protection against corrosion, protecting the surroundings against radiation, and protecting the glass cylinders from mechanical damage resulting from the pressure at the bottom of the disposal hole, and the modest generation of heat in the waste at the time of disposal resulting in a maximum temperature increase in the salt close to the waste of approx. 40 deg. C. These characteristics proved to considerably improve the safety margin with respect to unforeseen circumstances. The character of the salt dome and of the salt in the proposed disposal area offers in itself good protection against contact with the ground water outside the dome. The relatively large depth of 1200 and 2500 m of the salt surface also means that neither dome nor disposal facility will be appreciably influenced by glaciations or earthquakes. The chalk above the proposed disposal area is very tight and to retain radioactive matter effectively even in the precence of high concentrations of NaCL. The safety investigations included a number of natural processes and probable events such as the segregation of crystal water from overlooked salt minerals, faulty sealings of disposal holes, permeable fault zones in the chalk overlying the dome, the risk in connection with human penetration into the dome. These conditions will neither lead to the destruction of the waste casks or to the release of waste from the dome. Leaching of a cavern is the only situation which proved to result in a release of radioactive material to the biosphere, but the resulting doses was found to be small

Characteristics of solidified products containing radioactive molten salt waste.

Science.gov (United States)

Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Kim, Joon-Hyung

2007-11-01

The molten salt waste from a pyroprocess to recover uranium and transuranic elements is one of the problematic radioactive wastes to be solidified into a durable wasteform for its final disposal. By using a novel method, named as the GRSS (gel-route stabilization/solidification) method, a molten salt waste was treated to produce a unique wasteform. A borosilicate glass as a chemical binder dissolves the silicate compounds in the gel products to produce one amorphous phase while most of the phosphates are encapsulated by the vitrified phase. Also, Cs in the gel product is preferentially situated in the silicate phase, and it is vitrified into a glassy phase after a heat treatment. The Sr-containing phase is mainly phosphate compounds and encapsulated by the glassy phase. These phenomena could be identified by the static and dynamic leaching test that revealed a high leach resistance of radionuclides. The leach rates were about 10(-3) - 10(-2) g/m2 x day for Cs and 10(-4) - 10(-3) g/m2 x day for Sr, and the leached fractions of them were predicted to be 0.89% and 0.39% at 900 days, respectively. This paper describes the characteristics of a unique wasteform containing a molten salt waste and provides important information on a newly developed immobilization technology for salt wastes, the GRSS method.

Problems of the final storage of radioactive waste in salt formations

International Nuclear Information System (INIS)

Hofrichter, E.

1977-01-01

The geological conditions for the final storage of radioactive waste, the occurrence of salt formations, and the tectonics of salt domes are discussed. The safety of salt rocks, the impermeability of the rocks, and the thermal problems in the storage of high-activity waste are dealt with. Possibilities and preconditions of final storage in West Germany are discussed. (HPH) [de

Thermal denitration of high concentration nitrate salts waste water

International Nuclear Information System (INIS)

Hwang, D. S.; Oh, J. H.; Choi, Y. D.; Hwang, S. T.; Park, J. H.; Latge, C.

2003-01-01

This study investigated the thermodynamic and the thermal decomposition properties of high concentration nitrate salts waste water for the lagoon sludge treatment. The thermodynamic property was carried out by COACH and GEMINI II based on the composition of nitrate salts waste water. The thermal decomposition property was carried out by TG-DTA and XRD. Ammonium nitrate and sodium nitrate were decomposed at 250 .deg. C and 730 . deg. C, respectively. Sodium nitrate could be decomposed at 450 .deg. C in the case of adding alumina for converting unstable Na 2 O into stable Na 2 O.Al 2 O 3 . The flow sheet for nitrate salts waste water treatment was proposed based on the these properties data. These will be used by the basic data of the process simulation

Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

International Nuclear Information System (INIS)

Koyama, Tadafumi.

1994-01-01

A method is described for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities

Assessment of crushed salt consolidation and fracture healing processes in a nuclear waste repository in salt

International Nuclear Information System (INIS)

1984-11-01

For a nuclear waste repository in salt, two aspects of salt behavior are expected to contribute to favorable conditions for waste isolation. First, consolidation of crushed salt backfill due to creep closure of the underground openings may result in a backfill barrier with low permeability. Second, fractures created in the salt by excavation may heal under the influence of stress and temperature following sealing. This report reviews the status of knowledge regarding crushed salt consolidation and fracture healing, provides analyses which predict the rates at which the processes will occur under repository conditions, and develops requirements for future study. Analyses of the rate at which crushed salt will consolidate are found to be uncertain because of unexplained wide variation in the creep properties of crushed salt obtained from laboratory testing, and because of uncertainties in predictions of long term closure rates of openings in salt. This uncertainty could be resolved to a large degree by additional laboratory testing of crushed salt. Similarly, additional testing of fracture healing processes is required to confirm that healing will be effective under repository conditions. Extensive references, 27 figures, 5 tables

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

Energy Technology Data Exchange (ETDEWEB)

Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-08-08

An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

Engineering Options Assessment Report. Nitrate Salt Waste Stream Processing

Energy Technology Data Exchange (ETDEWEB)

Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-13

This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 above-ground UNS, and 79 candidate below-ground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing

Energy Technology Data Exchange (ETDEWEB)

Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-18

This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 aboveground UNS, and 79 candidate belowground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

International Nuclear Information System (INIS)

Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G.

2005-08-01

Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research

Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

Energy Technology Data Exchange (ETDEWEB)

Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G

2005-08-01

Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research.

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

International Nuclear Information System (INIS)

Akram, Najib

1993-01-01

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

Investigation of thermolytic hydrogen generation rate of tank farm simulated and actual waste

Energy Technology Data Exchange (ETDEWEB)

Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Howe, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-11-15

To support resolution of Potential Inadequacies in the Safety Analysis for the Savannah River Site (SRS) Tank Farm, Savannah River National Laboratory conducted research to determine the thermolytic hydrogen generation rate (HGR) with simulated and actual waste. Gas chromatography methods were developed and used with air-purged flow systems to quantify hydrogen generation from heated simulated and actual waste at rates applicable to the Tank Farm Documented Safety Analysis (DSA). Initial simulant tests with a simple salt solution plus sodium glycolate demonstrated the behavior of the test apparatus by replicating known HGR kinetics. Additional simulant tests with the simple salt solution excluding organics apart from contaminants provided measurement of the detection and quantification limits for the apparatus with respect to hydrogen generation. Testing included a measurement of HGR on actual SRS tank waste from Tank 38. A final series of measurements examined HGR for a simulant with the most common SRS Tank Farm organics at temperatures up to 140 °C. The following conclusions result from this testing.

Costs for off-site disposal of nonhazardous oil field wastes: Salt caverns versus other disposal methods

Energy Technology Data Exchange (ETDEWEB)

Veil, J.A.

1997-09-01

According to an American Petroleum Institute production waste survey reported on by P.G. Wakim in 1987 and 1988, the exploration and production segment of the US oil and gas industry generated more than 360 million barrels (bbl) of drilling wastes, more than 20 billion bbl of produced water, and nearly 12 million bbl of associated wastes in 1985. Current exploration and production activities are believed to be generating comparable quantities of these oil field wastes. Wakim estimates that 28% of drilling wastes, less than 2% of produced water, and 52% of associated wastes are disposed of in off-site commercial facilities. In recent years, interest in disposing of oil field wastes in solution-mined salt caverns has been growing. This report provides information on the availability of commercial disposal companies in oil-and gas-producing states, the treatment and disposal methods they employ, and the amounts they charge. It also compares cavern disposal costs with the costs of other forms of waste disposal.

Experimental results on salt concrete for barrier elements made of salt concrete in a repository for radioactive waste in a salt mine

International Nuclear Information System (INIS)

Gutsch, Alex-W.; Preuss, Juergen; Mauke, Ralf

2012-01-01

The Bartensleben rock salt mine in Germany was used as a repository for low and intermediate level radioactive waste from 1971 to 1991 and from 1994 to 1998. The repository with an overall volume of about 6 million m 3 has to be closed. Salt concrete is used for the refill of the voids of the repository. The concrete mixtures contain crushed salt instead of natural aggregates as the void filling material should be as similar to the salt rock as possible. Very high requirements regarding low heat development and little or even no cracking during concrete hardening had to be fulfilled even for the barrier elements made from salt concrete which separate the radioactive waste from the environment. Requirements for the salt concrete were set up with regard to the fluidity of the fresh concrete during the hardening process and its durability. In the view of a comprehensive numerical calculations of the temperature development and thermal stresses in the massive salt concrete elements of the backfill of the voids, experimental results for material properties of the salt concrete are presented: mixture of the salt concrete, thermodynamic properties (adiabatic heat release, thermal dilatation, thermal conductivity and heat capacity), mechanical short term properties, creep (under tension, under compression), autogenous shrinkage

Waste isolation facility description: bedded salt

Energy Technology Data Exchange (ETDEWEB)

1976-09-01

The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria. (LK)

Waste isolation facility description: bedded salt

International Nuclear Information System (INIS)

1976-09-01

The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria

Different Methods for Conditioning Chloride Salt Wastes

International Nuclear Information System (INIS)

De Angelis, G.; Fedeli, C.; Capone, M.; Marzo, G.A.; Mariani, M.; Da Ros, M.; Giacobbo, F.; Macerata, E.; Giola, M.

2015-01-01

Three different methods have been used to condition chloride salt wastes coming from pyro-processes. Two of them allow to synthesise sodalite, a naturally occurring mineral containing chlorine: the former, starting from Zeolite 4A, which transforms the zeolite into sodalite; the latter, which starts from kaolinite, giving sodalite as well. In addition, a new matrix, termed SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ), has been synthesised. It is able to form different mineral phases which occlude fission metals. The products from the different processes have been fully characterised. In particular the chemical durability of the final waste forms has been determined using the standard product consistency test. According to the results obtained, SAP seems to be a promising matrix for the incorporation of chloride salt wastes from pyro-processes. Financial support from the Nuclear Fission Safety Programme of the European Union (projects ACSEPT, contract FP7-CP-2007- 211 267, and SACSESS, Collaborative Project 323282), as well as from Italian Ministry for Economic Development (Accordo di Programma: Piano Annuale di Realizzazione 2008-2009) is gratefully acknowledged. (authors)

Status and integration of studies of gas generation in Hanford wastes

International Nuclear Information System (INIS)

Pederson, L.R.; Bryan, S.A.

1996-10-01

The purpose of this report is to review recent progress in determining the mechanism, kinetics, and stoichiometry of gas generation in Hanford waste tanks. Information has been gathered from the results of (1) laboratory studies with simulated wastes; (2) laboratory studies with actual waste core samples (Tanks SY-101 and SY-103); (3) studies of thermal and radiolytic reactions in the gas phase; (4) gas solubility evaluations; and (5) in-tank gas composition data. The results of laboratory studies using simulated wastes, which were aimed at determining chemical mechanisms responsible for gas generation, are summarized in Section 2. Emphasized are findings from work performed at the Georgia Institute of Technology (GIT), which was conducted under subcontract to Pacific Northwest National Laboratory (PNNL) and completed in FY 1996. Thermally activated pathways for the decomposition of hydroxyethylethylene-diaminetriacetic acid (HEDTA, trisodium salt) in simulated wastes were established by this work, among other accomplishments

Status and integration of studies of gas generation in Hanford wastes

Energy Technology Data Exchange (ETDEWEB)

Pederson, L.R.; Bryan, S.A.

1996-10-01

The purpose of this report is to review recent progress in determining the mechanism, kinetics, and stoichiometry of gas generation in Hanford waste tanks. Information has been gathered from the results of (1) laboratory studies with simulated wastes; (2) laboratory studies with actual waste core samples (Tanks SY-101 and SY-103); (3) studies of thermal and radiolytic reactions in the gas phase; (4) gas solubility evaluations; and (5) in-tank gas composition data. The results of laboratory studies using simulated wastes, which were aimed at determining chemical mechanisms responsible for gas generation, are summarized in Section 2. Emphasized are findings from work performed at the Georgia Institute of Technology (GIT), which was conducted under subcontract to Pacific Northwest National Laboratory (PNNL) and completed in FY 1996. Thermally activated pathways for the decomposition of hydroxyethylethylene-diaminetriacetic acid (HEDTA, trisodium salt) in simulated wastes were established by this work, among other accomplishments.

Pyrolytic conversion of plastic and rubber waste to hydrocarbons with basic salt catalysts

Science.gov (United States)

Wingfield, Jr., Robert C.; Braslaw, Jacob; Gealer, Roy L.

1985-01-01

The invention relates to a process for improving the pyrolytic conversion of waste selected from rubber and plastic to low molecular weight olefinic materials by employing basis salt catalysts in the waste mixture. The salts comprise alkali or alkaline earth compounds, particularly sodium carbonate, in an amount of greater than about 1 weight percent based on the waste feed.

Potential for creation of a salt dome following disposal of radioactive waste in a salt layer

International Nuclear Information System (INIS)

Fries, G.

1987-01-01

The study aims at quantifying the possibility of creation of a salt dome from a salt layer in which heat-emitting radioactive waste would be buried. Volume 1 describes the results of numerical computer simulations, and of laboratory-scale models in centrifuges. Volume 2 envisages, in a geological perspective, the origin of salt domes, the mechanisms of thei formation, and the associated parameters [fr

Potential for creation of a salt dome following disposal of radioactive waste in a salt layer

International Nuclear Information System (INIS)

Charo, L.; Habib, P.

1987-01-01

The study aims at quantifying the possibility of creation of a salt dome from a salt layer in which heat-emitting radioactive waste would be buried. Volume 1 describes the results of numerical computer simulations, and of laboratory-scale models in centrifuges. Volume 2 envisages, in a geological perspective, the origin of salt domes, the mechanisms of their formation, and the associated parameters [fr

Waste salt disposal at the Savannah River Plant

International Nuclear Information System (INIS)

Langton, C.A.; Oblath, S.B.; Pepper, D.W.; Wilhite, E.L.

1986-01-01

Waste salt solution, produced during processing of high-level nuclear waste, will be incorporated in a cement matrix for emplacement in an engineered disposal facility. Wasteform characteristics and disposal facility details will be presented along with results of a field test of wasteform contaminant release and of modeling studies to predict releases. 5 refs., 11 figs., 5 tabs

Waste package for a repository located in salt

International Nuclear Information System (INIS)

Basham, S.J. Jr.

1983-01-01

This paper describes the current status of the waste package designs for salt repositories. The status of the supporting studies of environment definition, corrosion of containment materials, and leaching of waste forms is also presented. Emphasis is on the results obtained in FY 83 and the planned effort in FY 84. 8 references, 3 figures, 1 table

Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

International Nuclear Information System (INIS)

Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

2004-01-01

Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Immobilization of IFR salt wastes in mortar

International Nuclear Information System (INIS)

Fisher, D.F.; Johnson, T.R.

1988-01-01

Portland cement-base mortars are being considered for immobilizing chloride salt wastes from the fuel cycle of an integral fast reactor (IFR). The IFR is a sodium-cooled fast reactor with metal fuel. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500 degrees C. This cell has a cadmium anode and a liquid salt electrolyte. The salt will be a low-melting mixture of alkaline and alkaline earth chlorides. This paper discusses one method being considered for immobilizing this treated salt, to disperse it in a portland cement-base motar, which would then be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canisters where it will solidify into a strong, leach-resistant material

Cerebral salt wasting following tuberculous meningoencephalitis in an infant

Directory of Open Access Journals (Sweden)

Syed Ahmed Zaki

2012-01-01

Full Text Available In patients with central nervous system disease, life-threatening hyponatremia can result from either the syndrome of inappropriate secretion of antidiuretic hormone or cerebral salt wasting. Clinical manifestations of the two conditions may be similar, but their pathogeneses and management protocols are different. Cerebral salt wasting syndrome is a disorder in which excessive natriuresis and hyponatremia occurs in patients with intracranial diseases. We report a 6-month-old girl with CSWS associated with tuberculous meningoencephalitis. She was diagnosed as having CSWS on the basis of hypovolemia, polyuria, natriuresis, and the relatively high level of fractional excretion of uric acid. Aggressive replacement of urine salt and water losses using 0.9% or 3% sodium chloride was done. Fludrocortisone was started at 0.1 mg twice daily on the seventh day of admission and was continued for 17 days.

Volume reduction of waste contaminated by fission product elements and plutonium using molten salt combustion

International Nuclear Information System (INIS)

McKenzie, D.E.; Grantham, L.F.; Paulson, R.B.

1979-01-01

In the Molten Salt Combustion Process, transuranic or β-γ organic waste and air are continuously introduced beneath the surface of a sodium carbonate-containing melt at a temperature of about 800 0 C. Complete combustion of the organic material to carbon dioxide and steam occurs without the conversion of nitrogen to nitrogen oxides. The noxious gases formed by combustion of the chloride, sulfur or phosphorus content of the waste instantly react with the melt to form the corresponding sodium compounds. These compounds as well as the ash and radionuclides are retained in the molten salt. The spent salt is either fused cast into an engineered disposal container or processed to recover salt and plutonium. Molten salt combustion reduces the waste to about 2% of its original volume. Many reactor or reprocessing wastes which cannot be incinerated without difficulty are readily combusted in the molten salt. A 50 kg/hr molten salt combustion system is being designed for the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. Construction of the combustor started during 1977, and combustor startup was scheduled for the spring of 1978

An experimental study on Sodalite and SAP matrices for immobilization of spent chloride salt waste

Science.gov (United States)

Giacobbo, Francesca; Da Ros, Mirko; Macerata, Elena; Mariani, Mario; Giola, Marco; De Angelis, Giorgio; Capone, Mauro; Fedeli, Carlo

2018-02-01

In the frame of Generation IV reactors a renewed interest in pyro-processing of spent nuclear fuel is underway. Molten chloride salt waste arising from the recovering of uranium and plutonium through pyro-processing is one of the problematic wastes for direct application of vitrification or ceramization. In this work, Sodalite and SAP have been evaluated and compared as potential matrices for confinement of spent chloride salt waste coming from pyro-processing. To this aim Sodalite and SAP were synthesized both in pure form and mixed with different glass matrices, i.e. commercially available glass frit and borosilicate glass. The confining matrices were loaded with mixed chloride salts to study their retention capacities with respect to the elements of interest. The matrices were characterized and leached for contact times up to 150 days at room temperature and at 90 °C. SEM analyses were also performed in order to compare the matrix surface before and after leaching. Leaching results are discussed and compared in terms of normalized releases with similar results reported in literature. According to this comparative study the SAP matrix with glass frit binder resulted in the best matrix among the ones studied, with respect to retention capacities for both matrix and spent fuel elements.

Destruction of high explosives and wastes containing high explosives using the molten salt destruction process

International Nuclear Information System (INIS)

Upadhye, R.S.; Brummond, W.A.; Pruneda, C.O.

1992-01-01

This paper reports the Molten Salt Destruction (MSD) Process which has been demonstrated for the destruction of HE and HE-containing wastes. MSD has been used by Rockwell International and by Anti-Pollution Systems to destroy hazardous wastes. MSD converts the organic constituents (including the HE) of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. In the case of HE-containing mixed wastes, any actinides in the waste are retained in the molten salt, thus converting the mixed wastes into low-level wastes. (Even though the MSD process is applicable to mixed wastes, this paper will emphasize HE-treatment.) The destruction of HE is accomplished by introducing it, together with oxidant gases, into a crucible containing a molten salt, such as sodium carbonate, or a suitable mixture of the carbonates of sodium, potassium, lithium and calcium. The temperature of the molten salt can be between 400 to 900 degrees C. The combustible organic components of the waste react with oxygen to produce carbon dioxide, nitrogen and steam

Salt-occluded zeolite waste forms: Crystal structures and transformability

International Nuclear Information System (INIS)

Richardson, J.W. Jr.

1996-01-01

Neutron diffraction studies of salt-occluded zeolite and zeolite/glass composite samples, simulating nuclear waste forms loaded with fission products, have revealed complex structures, with cations assuming the dual roles of charge compensation and occlusion (cluster formation). These clusters roughly fill the 6--8 angstrom diameter pores of the zeolites. Samples are prepared by equilibrating zeolite-A with complex molten Li, K, Cs, Sr, Ba, Y chloride salts, with compositions representative of anticipated waste systems. Samples prepared using zeolite 4A (which contains exclusively sodium cations) as starting material are observed to transform to sodalite, a denser aluminosilicate framework structure, while those prepared using zeolite 5A (sodium and calcium ions) more readily retain the zeolite-A structure. Because the sodalite framework pores are much smaller than those of zeolite-A, clusters are smaller and more rigorously confined, with a correspondingly lower capacity for waste containment. Details of the sodalite structures resulting from transformation of zeolite-A depend upon the precise composition of the original mixture. The enhanced resistance of salt-occluded zeolites prepared from zeolite 5A to sodalite transformation is thought to be related to differences in the complex chloride clusters present in these zeolite mixtures. Data relating processing conditions to resulting zeolite composition and structure can be used in the selection of processing parameters which lead to optimal waste forms

Electrodialysis-ion exchange for the separation of dissolved salts

International Nuclear Information System (INIS)

Baroch, C.J.; Grant, P.J.

1995-01-01

The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. This report describes the process of electrodialysis-ion exchange (EDIX) for treating aqueous wastes streams consisting of nitrates, sodium, organics, heavy metals, and radioactive species

Nuclear waste in sea or salt? No, wrong

International Nuclear Information System (INIS)

Damveld, H.; Van Duin, S.; Bannink, D.

1994-04-01

Eighteen years of successful action against ocean dumping and storage of nuclear waste in salt domes are reviewed for the Dutch situation. The aim of this book is to hand some support to those who want to act against trial borings, in particular the people living close to the most important salt domes in the Netherlands: Ternaard, Zuidwending, Pieterburen, Onstwedde, Winschoten, Schoonlo and Gasselte-Drouwen. In 1976 the Interdepartmental Commission on Nuclear Energy with its subcommission Radioactive Substances (ICK-RAS) was installed, along with a number of working groups, responsible for research. From 1978 onwards ocean dumping operations were accompanied by blockades and legal procedures, which led to a situation of the last dumping in 1982. The Dutch government then focused on nuclear waste storage in salt domes for which the OPLA research program was started. OPLA is the Dutch abbreviation for Storage on Land. The final report (phase 1 and 1a) of OPLA was published on 15 October 1993 as annex to the Dossier Nuclear Energy of the Dutch government. It has been decided that phase 1a is not followed by trial drillings, as planned before. Some critical remarks are made regarding the rounds of public participation and the notion of permanent retrievability of stored nuclear waste. Extensive use has been made of documentation from the Dutch government and parliament, and other literature and information sources

Temperature distributions in a salt formation used for the ultimate disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Ploumen, P.

1980-01-01

In the Federal Republic of Germany the works on waste disposal is focussed on the utilization of a salt formation for ultimate disposal of radioactive wastes. Heat released from the high-level waste will be dissipated in the salt and the surrounding geologic formations. The occuring temperature distributions will be calculated with computer codes. A survey of the developed computer codes will be shown; the results for a selected example, taking into account the loading sequence of the waste, the mine ventilation as well as an air gap between the waste and the salt, will be discussed. Furthermore it will be shown that by varying the disposal parameters, the maximum salt temperature can be below any described value. (Auth.)

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

International Nuclear Information System (INIS)

Wishau, R.

1998-05-01

Molten salt oxidation (MSO) is proposed as a 238 Pu waste treatment technology that should be developed for volume reduction and recovery of 238 Pu and as an alternative to the transport and permanent disposal of 238 Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious 238 Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of 238 Pu contaminated wastes is reduced to 30 drums. Further 238 Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious 238 Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose 238 Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment

Radiological consequences associated with human intrusion into radioactive waste repositories in salt formations

International Nuclear Information System (INIS)

Jacquier, P.

1989-01-01

The assessment of the radiological impact of human intrusion scenarios is extremely important in the case of repositories located in salt formations, since salt is obviously a valuable economic resource. Salt formations also represent a suitable medium for mining storage caverns for oil and gas. The scenario considered in this report is that of solution mining in salt formations to produce salt for human consumption. It is postulated that the salt is extracted by excavating a cavern through solution-mining and that, in the course of cavern enlargement, the waste is intercepted and drops to the bottom of the cavern. We have assumed that the intrusion takes place 500 or even 2 500 years after the repository has been sealed. The cases considered involve high-level vitrified waste or cemented alpha waste. The paper describes the assumptions on which the scenario is based and uses a simplified model to assess the radiological consequences associated with the ingestion of contaminated salt. The paper also provides details of a sensitivity/uncertainty analysis which identified several areas in which experimental studies should be either initiated or continued [fr

Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 16. Repository preconceptual design studies: BPNL waste forms in salt

International Nuclear Information System (INIS)

1978-04-01

This volume, Volume 16, ''Repository Preconceptual Design Studies: BPNL Waste Forms in Salt,'' is one of a 23 volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provide a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The waste forms assumed to arrive at the repository were supplied by Battelle Pacific Northwest Laboratories (BPNL). The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/17, ''Drawings for Repository Preconceptual Design Studies: BPNL Waste Forms in Salt.''

Modeling of waste/near field interactions for a waste repository in bedded salt: the Dynamic Network (DNET) model

International Nuclear Information System (INIS)

Cranwell, R.M.

1983-01-01

The Fuel Cycle Risk Analysis Division of Sandia National Laboratories has been funded by the US Nuclear Regulatory Commission to develop a methodology for use in assessing the long-term risk from the disposal of radioactive wastes in deep geologic formations. As part of this program, the Dynamic Network (DNET) model was developed to investigate waste/near field interactions associated with the disposal of radioactive wastes in bedded salt formations. The model is a quasi-multi-dimensional network model with capabilities for simulating processes such as fluid flow, heat transport, salt dissolution, salt creep, and the effects of thermal expansion and subsedence on the rock units surrounding the repository. The use of DNET has been demonstrated in the analysis of a hypothetical disposal site containing a bedded salt formation as the host medium for the repository. An example of this demonstration analysis is discussed. Furthermore, the outcome of sensitivity analyses performed on the DNET model are presented

Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.4

International Nuclear Information System (INIS)

1981-01-01

The present report deals with construction, operation and sealing of disposal facilities for high-level waste in a salt dome. It is volume 4 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. The safety investigations were carried out for a deep-hole disposal facility located in the salt dome on Mors. In principle the results of the investigations also apply to a shaft/mine disposal facility. The facility is designed for the disposal of vitrified high-level waste in the shape of glass canisters. There is a low concentration of waste in each canister, approx. 10%. Furthermore, it was selected to place the waste in an intermediate storage for about 40 years prior to its final disposal. Consequently, heat generation in the waste at the time of final disposal will be modest, resulting in low temperature increase in the salt. As an example, the highest temperature increase will be approx. 40 deg. C. and it will occur at the edge of the hole five years after disposal has taken place. Prior to disposal, the glass canisters are encased in steel casks with 15 cm thick walls. Three canisters are placed in each cask, and 215 casks are stacked on top on one another in each deep-hole from a depth of 1200 m to 2500 m underground. The additional encasing is designed to protect the glass from dissolution should any brine reach the disposal facility. Furthermore, the steel cask protects the glass canisters against pressure from the wall of the hole. The technical design of the disposal facility gives it a considerable safety margin against unexpected events. The investigations proved Cretaceous strata to constitute an effective secondary barrier that would prevent radioactive matter from travelling from the underlying disposal facility to the biosphere. (BP)

Chemical modeling of nuclear waste repositories in the salt repository project

International Nuclear Information System (INIS)

Jansen, G.; Raines, G.E.; Kircher, J.F.; Hubbard, N.

1985-01-01

Salt deposits contain small amounts of water as brine in fluid inclusions in halite and in hydrous minerals, e.g., clays, kieserite (MgSO 4 . H 2 O) and carnallite (KMgCl 3 . 6H 2 O). For the candidate salt deposits, the total amounts of water as volume % brine are: Palo Duro Basin, Texas, approximately 1.8; Paradox Basin, Utah, approximately 5.0 for the carnallite-marker zone, and less than approximately 0.5 below this zone; Gulf Coast salt domes, less than 0.15. For the Palo Duro and Paradox salt, the brines are Mg-rich (approximately 20,000 mg/L to approximately 100,000 mg/L) and sometimes Ca-rich (up to about 20,000 mg/L) NaCl brines. Brine migration calculations have been made using calculations of the time-variant thermal gradient around the waste packages and conservatively high brine volumes in the salt (5.0 volume % for the Texas and Utah sites and 0.5 volume % for the Gulf Coast) as input data. The maximum amounts of brine that eventually migrate to each waste package are about 1.0m 3 (for 5.0 volume % brine) and 0.2m 3 (for 0.5 volume % brine). With current conceptual designs for waste package overpacks (10 to 15 cm thick low-carbon steel), the waste package is not breached by uniform corrosion within 10,000 years. In brines this material thus far shows only uniform corrosion. For the expected conditions, where the brine is provided solely by brine migration, the brine is consumed by reaction with the iron of the overpack nearly as fast as it migrates to the waste package. Therefore, for the expected conditions, data about corrosion rates, radiolysis, etc., are not important. However, it is essential that accurate volumes of in-migrating brine can be calculated

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

International Nuclear Information System (INIS)

1984-01-01

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

Study on application of molten salt oxidation technology (MSO) for PVC wastes treatment

International Nuclear Information System (INIS)

Tran Thu Ha; Nguyen Hong Quy; Pham Quoc Ky; Nguyen Quang Long; Vuong Thu Bac; Dang Duc Nhan

2007-01-01

The project 'Study on application of molten salt oxidation (MSO) for PVC plastic wastes treatment' aims at three followings: 1) Installation of lab-scale MSO unit with essential compositions builds up foundation for the 2) estimation of waste destruction efficiency of the technology. 3) Based on the results of testing PVC - the chlorinated organic wastes on the lab-scale unit, the ability of the technology application at pilot-scale level will be primary estimated. The adjustment and correction of some compositions in the lab-scale unit theoretically designed during experiment overcame the shortages by design and fabrication such as heat distribution regime, feeding wastes and draining spent salt. These solutions adapt to the technical requirement of operation as well as scientific requirement of the research on MSO process. PVC waste treatment was tested on the MSO lab-scale unit in different conditions of operation temperature, superficial air velocity related to air/oxygen feeding rate, waste feeding rate. The testing results showed that destruction efficiency of chlorine in MSO technology was almost absolute. HCl and Cl 2 emission were insignificant in different operation conditions. HCl and Cl 2 emission depend on resident time and nature of molten salt. However, with inherent attributes of MSO technology emission of CO is not avoided in processing waste treatment. Therefore, finding active solutions for reduction CO emission is essential to complete the technology. The experiments also were carried in conditions of single molten salt (Na 2 CO 3 ) and molten (Na 2 CO 3 - K 2 CO 3 ) eutectic. The comparison of efficiency of these tests gives idea of using molten salt eutectic to reduce operation cost in MSO technology. Based on operation parameters and scientific verification results during experiments, the introductory procedure of waste treatment by MSO process was built up. Thereby, primary estimation of development of the technology in pilot-scale is given

Heat transfer analysis of the waste-container sleeve/salt configuration

International Nuclear Information System (INIS)

Callahan, G.D.; Ratigan, J.L.; Russell, J.E.; Fossum, A.F.

1975-01-01

Prior to this investigation, the heat transport considered was only that of straight conduction. The waste container, air gap, and sleeve arrangement was considered to be a single, consistent, time-dependent, heat-generating unit in intimate contact with the salt. The conduction model does not accurately model the heat transfer mechanisms available. Thus radiation and combined radiation and convection must also be considered in the determination of the temperature field. As would be expected, the canister temperatures are higher for the case of radiation across the airgap than those that result from conduction when the canister is in intimate contact with the salt. For the radiation case, the canister temperatures rise rapidly to a temperature of approximately 1,140 0 F and maintain an almost steady state condition for one year whereafter the temperatures slowly decrease. The far field temperatures, near the pillar centerline, are essentially equivalent for all cases. As time proceeds, the far field temperatures of the conduction models are about 15% different

Electrodialysis-ion exchange for the separation of dissolved salts

Energy Technology Data Exchange (ETDEWEB)

Baroch, C.J. [Wastren, Inc., Westminster, CO (United States); Grant, P.J. [Wastren, Inc., Hummelstown, PA (United States)

1995-10-01

The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. There is considerable interest in developing processes that remove or destroy the nitrate wastes. Electrodialysis-Ion Exchange (EDIX) is a possible process that should be more cost effective in treating aqueous waste steams. This report describes the EDIX process.

Calculations on the development in space and time of the temperature field around a repository of medium and high active wastes in a salt formation

International Nuclear Information System (INIS)

Delisle, G.

1980-01-01

The concept of nuclear waste disposal of th of the Federal Republic of Germany calls for the burial of the wastes within a salt formation. A small portion of the wastes will generate heat after the disposal procedure. A temperature rise within the salt formation, in space and time limited, will be the consequence. The temperature change at any point in the near or far field of the disporal area can be calculated with the aid of numerical models. The thermal parameters representative for the bulk material of the Zechstein formation in NW-Germany, on which the calculations are based, will be discussed in detail. The interrelation between the concentration of heat producing wastes in the disposal field and the maximum average temperature in the salt formation will be treated. By defining numerical models, which are based on assumed shapes of a salt dome and a disposal area, the temperature development in the near and far field of a nuclear repository are shown. (orig.) [de

Backfill barriers for nuclear waste repositories in salt

Energy Technology Data Exchange (ETDEWEB)

Nowak, E J; Odoj, R; Merz, E [eds.

1981-06-01

Backfill materials were evaluated for containment of radionuclides, chemical modification of brine, and sensitivity to hydrothermal conditions. Experimental conditions were relevant to nuclear waste isolation in bedded salt. They were based on geologic conditions at the site of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, USA. Conclusions are: backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier; bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved ions, or the sorption properties of radionuclide species; effects of irradiation from high level waste should be investigated.

Using TOUGH2 to model the coupled effects of gas generation, repository consolidation, and multiphase brine and gas flow at the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Freeze, G.A.; Larson, K.W.; Davies, P.B.; Webb, S.W.

1995-01-01

The Waste Isolation Pilot Plant (WIPP) is a US Department of Energy facility designed to demonstrate the safe underground disposal of transuranic waste. Following waste emplacement, each room will be backfilled with crushed salt. Due to deviatoric stress introduced by excavation, the walls of the waste disposal rooms in the repository will deform over time, consolidating waste containers and salt backfill, thereby decreasing the void volume of the repository. Long-term repository assessment must consider the processes of gas generation, room closure and expansion due to salt creep, and multiphase (brine and gas) fluid flow, as well as the complex coupling between these three processes. Stone (1992) used the mechanical creep closure code SANCHO to simulate the closure of a single, perfectly sealed disposal room filled with waste and backfill. The results of the SANCHO f-series simulations provide a relationship between gas generation, room closure, and room pressure. Several methods for coupling this relationship with multiphase fluid flow into and out of a room were examined by Freeze et al. TOUGH2 was employed to couple the processes of gas generation, room closure/consolidation, and multiphase brine and gas flow. Two empirically-based methods for approximating salt creep and room consolidation were implemented in TOUGH2: the pressure-time-porosity line interpolation approach and the fluid-phase-salt approach. Both approaches utilized links to the SANCHO f-series simulation results to calculate room-void-volume changes with time during a simulation

An improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds

International Nuclear Information System (INIS)

Eun, H.C.; Cho, Y.Z.; Lee, T.K.; Kim, I.T.; Park, G.I.; Lee, H.S.

2013-01-01

In this paper, an improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds was performed to determine optimum operating conditions. It was very important to maintain the pressure in the distillation chamber below 10 Torr for a high efficiency (salt recovery >99 %) of the salt distillation. This required increasing the salt vaporization and condensation rates in the distillation system. It was confirmed that vaporization and condensation rates could be improved controlling the given temperature of top of the condensation chamber. In the distillation tests of the salt wastes containing rare earth compounds, the operation time at a given temperature was greatly reduced changing the given temperature of top of the condensation chamber from 780 to 700 deg C. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Wishau, R.

1998-05-01

Molten salt oxidation (MSO) is proposed as a {sup 238}Pu waste treatment technology that should be developed for volume reduction and recovery of {sup 238}Pu and as an alternative to the transport and permanent disposal of {sup 238}Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious {sup 238}Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of {sup 238}Pu contaminated wastes is reduced to 30 drums. Further {sup 238}Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious {sup 238}Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose {sup 238}Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment.

Rock salt as a medium for long-term isolation of radioactive wastes - a reassessment

International Nuclear Information System (INIS)

Chaturvedi, L.

1985-01-01

Rock salt has been regarded as a suitable medium for the permanent disposal of high and medium level radioactive wastes since the National Academy of Sciences recommended it in 1957. As a result of detained site-specific studies conducted for the Waste Isolation Pilot Plant (WIPP) project in New Mexico, however, several potential problems which are unique to bedded salt deposits have emerged. These include 1) the need to delineate the extent and rate of past dissolution and projections for the future, 2) the origin and significance of brines often found underlying the salt beds, 3) the rate and volume of migration of brine from the salt crystals towards the heat producing waste canisters, 4) the creep rates and implications for retrievability, and 5) the existence of potash and oil and gas resources with implications of human intrusion in the future. These questions will also be faced for sites in salt domes with added complications due to more complex structure and hydrology. The experience at WIPP shows that the site characterization process for high level waste repositories in bedded or dome salt should aim at identifying the important issues of site suitability early in the process and a clear program should be established to address these issues

Salt Repository Project waste emplacement mode decision paper: Revison 1

International Nuclear Information System (INIS)

1987-08-01

This paper provides a recommendation as to the mode of waste emplacement to be used as the current basis for site characterization activity for the Deaf Smith County, Texas, high level nuclear waste repository site. It also presents a plan for implementing the recommendation so as to provide a high level of confidence in the project's success. Since evaluations of high-level waste disposal in geologic repositories began in the 1950s, most studies emplacement in salt formations employed the vertical orientation for emplacing waste packages in boreholes in the floor of the underground facility. This orientation was used in trials at Project Salt Vault in the 1960s. The Waste Isolation Pilot Plant (WIPP) has recently settled on a combination of vertical and horizontal modes for various waste types. This paper analyzes the information available and develops a project position upon which to base current site characterization activities. The position recommended is that the SRP should continue to use the vertical waste emplacement mode as the reference design and to carry the horizontal mode as a ''passive'' alternative. This position was developed based upon the conclusions of a decision analysis, risk assessment, and cost/schedule impact assessment. 52 refs., 6 figs., 1 tab

Waste form dissolution in bedded salt

International Nuclear Information System (INIS)

Kaufman, A.M.

1980-01-01

A model was devised for waste dissolution in bedded salt, a hydrologically tight medium. For a typical Spent UnReprocessed Fuel (SURF) emplacement, the dissolution rate wll be diffusion limited and will rise to a steady state value after t/sub eq/ approx. = 250 (1+(1-epsilon 0 ) K/sub D//epsilon 0 ) (years) epsilon 0 is the overpack porosity and K/sub d/ is the overpack sorption coefficient. The steady state dissolution rate itself is dominated by the solubility of UO 2 . Steady state rates between 5 x 10 -5 and .5 (g/year) are achievable by SURF emplacements in bedded salt without overpack, and rates between 5 x 10 -7 and 5 x 10 -3 (g/year) with an overpack having porosity of 10 -2

Performance analysis of conceptual waste package designs in salt repositories

International Nuclear Information System (INIS)

Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

1984-01-01

A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

Long-term interactions of full-scale cemented waste simulates with salt brines

Energy Technology Data Exchange (ETDEWEB)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-07-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO{sub 3} solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

Long-term interactions of full-scale cemented waste simulates with salt brines

International Nuclear Information System (INIS)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-01-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO 3 solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

Dechlorination and Stabilization of Molten Salt Waste by Using xSiO2-yAl2O3- zP2O5 at Melting Temperature

International Nuclear Information System (INIS)

Park, Hwanseo; Kim, Intae; Kim, Hwanyoung; Kim, Joonhyung

2007-01-01

Molten salt waste, which is generated from the pyroprocess to separate uranium and trans-uranium elements from spent nuclear fuel, has been interested to researchers in the radioactive waste management. For its final disposal, direct immobilization into a suitable host matrix or indirect solidification by other chemical routes requires the control of chlorides and its volatility since molten salt wastes mainly consist of volatile metal chlorides. Glass-bonded sodalite (Na 6 M 2 Al 6 Si 6 O 24 Cl 2 , 1-5) suggested by Argonne National Laboratory (ANL), to the present, could be a practical solution to the immobilization of this waste, where waste form can be fabricated at about 915 .deg., lower than the melting temperature of many borosilicate glasses ( -1150 .deg.). A wet dechlorination to oxides or a thermal conversion into borate glass was suggested to remove Cl from salt waste (6-7) and it seemed that the preference of radionuclides for the intended chemical conversions or immobilizations described above could be hardly accomplished or failed, except the phosphate precipitation method suggested by Volkovich and his co-workers (8). Our research group suggested a novel method to treat molten salt waste, named GRSS (Gel-Route Stabilization/Solidification) using Si-P-Al system as a gel-forming system. This showed little vaporization during high temperature process and good leach resistance on Cs and Sr. As another method, this study suggested a method to stabilize molten salt wastes by using xSiO 2 -yAl 2 O 3 - zP 2 O 5 material. GRSS method is considered as a 'reaction system' to completely convert salt waste into stable product while the inorganic material used in this study is a stabilizer for salt wastes. Using this material, this study investigated the reactivity on different metal chlorides, thermal stability, leach-resistance and etc

Electrodialysis-based separation process for salt recovery and recycling from waste water

Science.gov (United States)

Tsai, Shih-Perng

1997-01-01

A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid.

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

International Nuclear Information System (INIS)

1984-01-01

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

Immobilization of LiCl-Li 2 O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

2018-04-01

In this study, salt occlusion and hydrothermal processes were used to make chlorosodalite through reaction with a high-LiCl salt simulating a waste stream following pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and aide in densification. Hydrothermal processes included reaction of the salt simulant in an acid digestion vessel with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

Waste package materials testing for a salt repository: 1983 status summary report

International Nuclear Information System (INIS)

Moak, D.P.

1986-09-01

The United States plans to safely dispose of nuclear waste in deep, stable geologic formations. As part of these plans, the US Department of Energy is sponsoring research on the designing and testing of waste packages and waste package materials. This fiscal year 1983 status report summarizes recent results of waste package materials testing in a salt environment. The results from these tests will be used by waste package designers and performance assessment experts. Release characteristics data are available on two waste forms (spent fuel and waste-containing glass) that were exposed to leaching tests at various radiation levels, temperatures, pH, glass surface area to solution volume ratios, and brine solutions simulating expected salt repository conditions. Candidate materials tested for corrosion resistance and other properties include iron alloys; TI-CODE 12, the most promising titanium alloy for containment; and nickel alloys. In component interaction testing, synergistic effects have not ruled out any candidate material. 21 refs., 37 figs., 15 tabs

Treatment Study Plan for Nitrate Salt Waste Remediation Revision 1.0

Energy Technology Data Exchange (ETDEWEB)

Juarez, Catherine L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vigil-Holterman, Luciana R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Naranjo, Felicia Danielle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-07

The two stabilization treatment methods that are to be examined for their effectiveness in the treatment of both the unremediated and remediated nitrate salt wastes include (1) the addition of zeolite and (2) cementation. Zeolite addition is proposed based on the results of several studies and analyses that specifically examined the effectiveness of this process for deactivating nitrate salts. Cementation is also being assessed because of its prevalence as an immobilization method used for similar wastes at numerous facilities around the DOE complex, including at Los Alamos. The results of this Treatment Study Plan will be used to provide the basis for a Resource Conservation and Recovery Act (RCRA) permit modification request of the LANL Hazardous Waste Facility Permit for approval by the New Mexico Environment Department-Hazardous Waste Bureau (NMED-HWB) of the proposed treatment process and the associated facilities.

Novel waste printed circuit board recycling process with molten salt

OpenAIRE

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450?470??C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, a...

Assessment of tectonic hazards to waste storage in interior-basin salt domes

International Nuclear Information System (INIS)

Kehle, R.

1979-01-01

Salt domes in the northern Gulf of Mexico may make ideal sites for storage of radioactive waste because the area is tectonically quiet. The stability of such salt domes and the tectonic activity are discussed

Low disposal of radioactive wastes in salt formations of the Federal Republic of Germaany

International Nuclear Information System (INIS)

Albrecht, E.

1980-01-01

The salt formations of northern Europe are generally suitable for the storage of radioactive wastes because the region is largely free from earthquakes and the salt formations known as diapires provide effective hydrological sealing. The Federal Republic of Germany employed the Asse Salt Mine of Lower Saxony for research in waste storage. More recently, exploratory work has begun on the construction of a large recycling and disposal plant at the Gorleben salt dome. The geology, hydrology, rock mechanics, and seismicity of the two sites are briefly discussed, including a discussion of experiences gained so far from the Asse site. 11 refs

A reactive distillation process for the treatment of LiCl-KCl eutectic waste salt containing rare earth chlorides

Energy Technology Data Exchange (ETDEWEB)

Eun, H.C., E-mail: ehc2004@kaeri.re.kr; Choi, J.H.; Kim, N.Y.; Lee, T.K.; Han, S.Y.; Lee, K.R.; Park, H.S.; Ahn, D.H.

2016-11-15

The pyrochemical process, which recovers useful resources (U/TRU metals) from used nuclear fuel using an electrochemical method, generates LiCl-KCl eutectic waste salt containing radioactive rare earth chlorides (RECl{sub 3}). It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste salt in a hot-cell facility. For this reason, a reactive distillation process using a chemical agent was achieved as a method to separate rare earths from the LiCl-KCl waste salt. Before conducting the reactive distillation, thermodynamic equilibrium behaviors of the reactions between rare earth (Nd, La, Ce, Pr) chlorides and the chemical agent (K{sub 2}CO{sub 3}) were predicted using software. The addition of the chemical agent was determined to separate the rare earth chlorides into an oxide form using these equilibrium results. In the reactive distillation test, the rare earth chlorides in LiCl-KCl eutectic salt were decontaminated at a decontamination factor (DF) of more than 5000, and were mainly converted into oxide (Nd{sub 2}O{sub 3}, CeO{sub 2}, La{sub 2}O{sub 3}, Pr{sub 2}O{sub 3}) or oxychloride (LaOCl, PrOCl) forms. The LiCl-KCl was purified into a form with a very low concentration (<1 ppm) for the rare earth chlorides.

Reconsolidated Salt as a Geotechnical Barrier

International Nuclear Information System (INIS)

Hansen, Francis D.; Gadbury, Casey

2015-01-01

Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt

Reconsolidated Salt as a Geotechnical Barrier

Energy Technology Data Exchange (ETDEWEB)

Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gadbury, Casey [USDOE Carlsbad Field Office, NM (United States)

2015-11-01

Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt

Thermal-gradient migration of brine inclusions in salt

International Nuclear Information System (INIS)

Yagnik, S.K.

1982-02-01

It has been proposed that the high level nuclear waste be buried deep underground in a suitable geologic formation. Natural salt deposits have been under active consideration as one of the geologic formations where a nuclear waste repository may be built in future. The salt deposits, however, are known to contain a small amount (about 0.5 vol.%) of water in the form of brine inclusions which are dispersed throughout the medium. The temperature gradients imposed by the heat generating nuclear waste will mobilize these brine inclusions. It is important to know the rate and the amount of brine accumulating at the waste packages to properly evaluate the performance of a nuclear waste repository. An extensive experimental investigation of the migration velocities of brine inclusions in synthetic single crystals of NaCl and in polycrystalline natural salt crystals has been conducted. The results show that in a salt repository the brine inclusions within a grain would move with the diffusion controlled velocities. The brine reaching a grain boundary may be swept across, if the thermal gradient is high enough. Grain boundaries in polycrystalline rock salt are apparently quite weak and open up due to drilling the hole for a waste canister and to the thermal stresses which accompany the thermal gradient produced by the heat generating waste. The enhanced porosity allows the water reaching the grain boundary to escape by a vapor transport process

Removal of salt from high-level waste tanks by density-driven circulation or mechanical agitation

International Nuclear Information System (INIS)

Kiser, D.L.

1981-01-01

Twenty-two high-level waste storage tanks at the Savannah River Plant are to be retired in the tank replacement/waste transfer program. The salt-removal portion of this program requires dissolution of about 19 million liters of salt cake. Steam circulation jets were originally proposed to dissolve the salt cake. However, the jets heated the waste tank to 80 to 90 0 C. This high temperature required a long cooldown period before transfer of the supernate by jet, and increased the risk of stress-corrosion cracking in these older tanks. A bench-scale investigation at the Savannah River Laboratory developed two alternatives to steam-jet circulation. One technique was density-driven circulation, which in bench tests dissolved salt at the same rate as a simulated steam circulation jet but at a lower temperature. The other technique was mechanical agitation, which dissolved the salt cake faster and required less fresh water than either density-driven circulation or the simulated steam circulation jet. Tests in an actual waste tank verified bench-scale results and demonstrated the superiority of mechanical agitation

Using Aspen simulation package to determine solubility of mixed salts in TRU waste evaporator bottoms

Energy Technology Data Exchange (ETDEWEB)

Hatchell, J.L.

1998-03-01

Nitric acid from plutonium process waste is a candidate for waste minimization by recycling. Process simulation software packages, such as Aspen, are valuable tools to estimate how effective recovery processes can be, however, constants in equations of state for many ionic components are not in their data libraries. One option is to combine single salt solubility`s in the Aspen model for mixed salt system. Single salt solubilities were regressed in Aspen within 0.82 weight percent of literature values. These were combined into a single Aspen model and used in the mixed salt studies. A simulated nitric acid waste containing mixed aluminum, calcium, iron, magnesium and sodium nitrate was tested to determine points of solubility between 25 and 100 C. Only four of the modeled experimental conditions, at 50 C and 75 C, produced a saturated solution. While experimental results indicate that sodium nitrate is the first salt to crystallize out, the Aspen computer model shows that the most insoluble salt, magnesium nitrate, the first salt to crystallize. Possible double salt formation is actually taking place under experimental conditions, which is not captured by the Aspen model.

Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

International Nuclear Information System (INIS)

Hull, A.B.; Williams, L.B.

1985-07-01

Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs

Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

Energy Technology Data Exchange (ETDEWEB)

Hull, A.B.; Williams, L.B.

1985-07-01

Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs.

Stabilization/Solidification of radioactive molten salt waste by using xSiO2-yAl2O3-zP2O5 material

International Nuclear Information System (INIS)

Hwan-Seo Park; In-Tae Kim; Yong-Zun Cho; Seong-Won Park; Eung-Ho Kim

2008-01-01

Molten salt waste generated from the electro metallurgical process to recover uranium and transuranic elements is considered as one of problematic wastes to be difficult to immobilize into a durable for final disposal. As an alternative, this study suggested a new method performed at molten state, where dechlorination was achieved with a new inorganic material containing SiO 2 , Al 2 O 3 and P 2 O 5 (SAP). The SAP as a reactive material to molten salt was prepared by a conventional sol-gel process. The prepared SAPs were reacted with each metal chloride, LiCl, CsCl, SrCl 2 and CeCl 3 at 650 deg. C for 6 hours and also were reacted with simulated salt waste consisting of 90 wt% LiCl, 6.8 wt% CsCl and 3.2 wt% SrCl 2 at different waste loading. All the reactions were carried out in oxidative atmosphere and metal chlorides were effectively converted into stable products under a reasonable reaction ratio

User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

International Nuclear Information System (INIS)

Lindner, E.N.; St John, C.M.; Hart, R.D.

1984-02-01

SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

Hydrological methods preferentially recover cesium from nuclear waste salt cake

International Nuclear Information System (INIS)

Brooke, J.N.; Hamm, L.L.

1997-01-01

The Savannah River Site is treating high level radioactive waste in the form of insoluble solids (sludge), crystallized salt (salt cake), and salt solutions. High costs and operational concerns have prompted DOE to look for ways to improve the salt cake treatment process. A numerical model was developed to evaluate the feasibility of pump and treat technology for extracting cesium from salt cake. A modified version of the VAM3DCG code was used to first establish a steady-state flow field, then to simulate 30 days of operation. Simulation results suggest that efficient cesium extraction can be obtained with low displacement volumes. The actual extraction process will probably be less impressive because of nonuniform properties. 2 refs., 2 figs

Risk assessment of nonhazardous oil-field waste disposal in salt caverns.

Energy Technology Data Exchange (ETDEWEB)

Elcock, D.

1998-03-05

In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes.

Risk assessment of nonhazardous oil-field waste disposal in salt caverns

International Nuclear Information System (INIS)

Elcock, D.

1998-01-01

In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes

Thermal decomposition of nitrate salts liquid waste for the lagoon sludge treatment

International Nuclear Information System (INIS)

Hwang, D. S.; Oh, J. H.; Kim, Y. K.; Lee, K. Y.; Choi, Y. D.; Hwang, S. T.; Park, J. H.

2004-01-01

This study investigated the thermal decomposition property of nitrate salts liquid waste which is produced in a series of the processes for the sludge treatment. Thermal decomposition property was analyzed by TG/DTA and XRD. Most ammonium nitrate in the nitrate salts liquid waste was decomposed at 250 .deg. C and calcium nitrate was decomposed and converted into calcium oxide at 550 .deg. C. Sodium nitrate was decomposed at 700 .deg. C and converted into sodium oxide which reacts with water easily. But sodium oxide was able to convert into a stable compound by adding alumina. Therefore, nitrate salts liquid waste can be treated by two steps as follows. First, ammonium nitrate is decomposed at 250 .deg. C. Second, alumina is added in residual solid sodium nitrate and calcium nitrate and these are decomposed at 900 .deg. C. Final residue consists of calcium oxide and Na 2 O.Al 2 O 3 and can be stored stably

Modeling of Sulfate Double-Salt in Nuclear Wastes

International Nuclear Information System (INIS)

Toghiani, B.; Lindner, J.S.; Weber, C.F.; Hunt, R.D.

2000-01-01

The Environmental Simulation Program (ESP) continues to adequately predict the solubility of most key chemical systems in the Hanford tank waste. For example, the ESP predictions were in fair agreement with the solubility experiments for the fluoride-phosphate system, although ESP probably underestimates the aqueous amounts. Due to the importance of this system in the formation of pipeline plugs, additional experiments have been made at elevated temperatures, and improvements to the ESP database will be made. ESP encountered problems with sulfate systems because the Public database for ESP does not include anhydrous sodium sulfate in mixed solutions below 32.4 C. This limitation leads to convergence problems and to spurious predictions of solubility near the transition point with sodium sulfate decahydrate when other salts such as sodium nitrate are present. However, ESP was able to make reasonable solubility predictions with a corrected database, demonstrating the need to validate and document the various databases that can be used by ESP. Even though ESP does not include the sulfate-nitrate double salt, this omission does not appear to be a major problem. The solubility predictions with and without the sulfate-nitrate double salt are comparable. In sharp contrast, the sulfate-fluoride double salt is included, but ESP still underestimates solubility in some cases. This problem can misrepresent the ionic strength of the solution, which is an important factor in the formation of pipeline plugs. Solubility tests on the sulfate-fluoride system are planned to provide additional data at higher temperatures and in caustic solutions. These results will be used to improve the range and accuracy of ESP predictions. ESP will continue to provide important predictions for waste processing operations while being evaluated and improved. For example, ESP will be used to determine the amount of water for the saltcake dissolution efforts at Hanford. When ESP underestimates the

Effects of heating on salt-occluded zeolite

International Nuclear Information System (INIS)

Lewis, M.A.; Hash, M.C.; Pereira, C.; Ackerman, J.P.

1996-01-01

The electrometallurgical treatment of spent nuclear fuel generates a waste stream of fission products in the electrolyte, LiCl-KCl eutectic salt. Argonne National Laboratory is developing a mineral waste form for this waste stream. The waste form consists of a composite formed by hot pressing salt-occluded zeolite and a glass binder. Pressing conditions must be judiciously chosen. For a given pressure, increasing temperatures and hold times give denser products but the zeolite is frequently converted to sodalite. Reducing the temperature or hold time leads to a porous zeolite composite. Therefore, conditions that affect the thermal stability of salt-occluded zeolite both with and without glass are being investigated in an ongoing study. The parameters varied in this stage of the work were heating time, temperature, salt loading, and glass content. The heat-treated samples were examined primarily by X-ray diffraction. Large variations were found in the rate at which salt-occluded zeolite converted to other phases such as nepheline, salt, and sodalite. The products depended on the initial salt loading. Heating times required for these transitions depended on the procedure and temperature used to prepare the salt-occluded zeolite. Mixtures of glass and zeolite reacted much faster than the pure salt-occluded zeolite and were almost always converted to sodalite

Prostaglandin-E2 Mediated Increase in Calcium and Phosphate Excretion in a Mouse Model of Distal Nephron Salt Wasting.

Directory of Open Access Journals (Sweden)

Manoocher Soleimani

Full Text Available Contribution of salt wasting and volume depletion to the pathogenesis of hypercalciuria and hyperphosphaturia is poorly understood. Pendrin/NCC double KO (pendrin/NCC-dKO mice display severe salt wasting under basal conditions and develop profound volume depletion, prerenal renal failure, and metabolic alkalosis and are growth retarded. Microscopic examination of the kidneys of pendrin/NCC-dKO mice revealed the presence of calcium phosphate deposits in the medullary collecting ducts, along with increased urinary calcium and phosphate excretion. Confirmatory studies revealed decreases in the expression levels of sodium phosphate transporter-2 isoforms a and c, increases in the expression of cytochrome p450 family 4a isotypes 12 a and b, as well as prostaglandin E synthase 1, and cyclooxygenases 1 and 2. Pendrin/NCC-dKO animals also had a significant increase in urinary prostaglandin E2 (PGE-2 and renal content of 20-hydroxyeicosatetraenoic acid (20-HETE levels. Pendrin/NCC-dKO animals exhibit reduced expression levels of the sodium/potassium/2chloride co-transporter 2 (NKCC2 in their medullary thick ascending limb. Further assessment of the renal expression of NKCC2 isoforms by quantitative real time PCR (qRT-PCR reveled that compared to WT mice, the expression of NKCC2 isotype F was significantly reduced in pendrin/NCC-dKO mice. Provision of a high salt diet to rectify volume depletion or inhibition of PGE-2 synthesis by indomethacin, but not inhibition of 20-HETE generation by HET0016, significantly improved hypercalciuria and salt wasting in pendrin/NCC dKO mice. Both high salt diet and indomethacin treatment also corrected the alterations in NKCC2 isotype expression in pendrin/NCC-dKO mice. We propose that severe salt wasting and volume depletion, irrespective of the primary originating nephron segment, can secondarily impair the reabsorption of salt and calcium in the thick ascending limb of Henle and/or proximal tubule, and reabsorption of

Leaching due to hygroscopic water uptake in cemented waste containing soluble salts

DEFF Research Database (Denmark)

Brodersen, K.

1992-01-01

conditions, condensation of water vapour will result in generation of a certain amount of liquid in the form of a strong salt solution. The volume of liquid may well exceed the storage capacity of the pore system in the cemented material and in the release of a limited amount of free contaminated solution......Considerable amounts of easily soluble salts such as sodium nitrate, sulphate, or carbonate are introduced into certain types of cemented waste. When such materials are stored in atmospheres with high relative humidity or disposed or by shallow land burial under unsaturated, but still humid....... A model of the quantitative aspects for the equilibrium situation is presented. Experiments with hygroscopic water uptake support the model and give indications about the rate of the process. The release mechanism is only thought to be important for radionuclides which are not fixed in a low...

Accelerator molten-salt breeder reactor

International Nuclear Information System (INIS)

Furukawa, Kazuo; Kuroi, Hideo; Kato, Yoshio; Oomichi, Toshihiko.

1979-01-01

Purpose: To obtain fission products and to transmute transuranium elements and other radioactive wastes by the use of Accelerator Molten-Salt Breeder Reactor. Constitution: Beams from an accelerator pipe at one end of a target vessel is injected through a window into target molten salts filled inside of the target vessel. The target molten salts are subjected to pump recycling or spontaneous convection while forcively cooled by blanket molten salts in an outer vessel. Then, energy is recovered from the blanket molten salts or the target molten salts at high temperatures through electric power generation or the like. Those salts containing such as thorium 232 and uranium 238 are used as the blanket molten salts so that fission products may be produced by neutrons generated in the target molten salts. PbCl 2 -PbF 2 and LiF-BeF 2 -ThF 4 can be used as the target molten salts and as the blanket molten salts respectively. (Seki, T.)

Brine migration in salt and its implications in the geologic disposal of nuclear waste

International Nuclear Information System (INIS)

Jenks, G.H.; Claiborne, H.C.

1981-12-01

This report respresents a comprehensive review and analysis of available information relating to brine migration in salt surrounding radioactive waste in a salt repository. The topics covered relate to (1) the characteristics of salt formations and waste packages pertinent to considerations of rates, amounts, and effects of brine migration, (2) experimental and theoretical information on brine migration, and (3) means of designing to minimize any adverse effects of brine migration. Flooding, brine pockets, and other topics were not considered, since these features will presumably be eliminated by appropriate site selection and repository design. 115 references

Options Assessment Report: Treatment of Nitrate Salt Waste at Los Alamos National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-12-17

This report documents the methodology used to select a method of treatment for the remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The method selected should treat the containerized waste in a manner that renders the waste safe and suitable for transport and final disposal in the Waste Isolation Pilot Plant (WIPP) repository, under specifications listed in the WIPP Waste Acceptance Criteria (DOE/CBFO, 2013). LANL recognizes that the results must be thoroughly vetted with the New Mexico Environment Department (NMED) and that a modification to the LANL Hazardous Waste Facility Permit is a necessary step before implementation of this or any treatment option. Likewise, facility readiness and safety basis approvals must be received from the Department of Energy (DOE). This report presents LANL’s preferred option, and the documentation of the process for reaching the recommended treatment option for RNS and UNS waste, and is presented for consideration by NMED and DOE.

Options assessment report: Treatment of nitrate salt waste at Los Alamos National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-09-16

This report documents the methodology used to select a method of treatment for the remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The method selected should treat the containerized waste in a manner that renders the waste safe and suitable for transport and final disposal in the Waste Isolation Pilot Plant (WIPP) repository, under specifications listed in the WIPP Waste Acceptance Criteria (DOE/CBFO, 2013). LANL recognized that the results must be thoroughly vetted with the New Mexico Environment Department (NMED) and the a modification to the LANL Hazardous Waste Facility Permit is a necessary step before implementation of this or any treatment option. Likewise, facility readiness and safety basis approvals must be received from the Department of Energy (DOE). This report presents LANL's preferred option, and the documentation of the process for reaching the recommended treatment option for RNS and UNS waste, and is presented for consideration by NMED and DOE.

Experiments and Modeling in Support of Generic Salt Repository Science

International Nuclear Information System (INIS)

Bourret, Suzanne Michelle; Stauffer, Philip H.; Weaver, Douglas James; Caporuscio, Florie Andre; Otto, Shawn; Boukhalfa, Hakim; Jordan, Amy B.; Chu, Shaoping; Zyvoloski, George Anthony; Johnson, Peter Jacob

2017-01-01

Salt is an attractive material for the disposition of heat generating nuclear waste (HGNW) because of its self-sealing, viscoplastic, and reconsolidation properties (Hansen and Leigh, 2012). The rate at which salt consolidates and the properties of the consolidated salt depend on the composition of the salt, including its content in accessory minerals and moisture, and the temperature under which consolidation occurs. Physicochemical processes, such as mineral hydration/dehydration salt dissolution and precipitation play a significant role in defining the rate of salt structure changes. Understanding the behavior of these complex processes is paramount when considering safe design for disposal of heat-generating nuclear waste (HGNW) in salt formations, so experimentation and modeling is underway to characterize these processes. This report presents experiments and simulations in support of the DOE-NE Used Fuel Disposition Campaign (UFDC) for development of drift-scale, in-situ field testing of HGNW in salt formations.

Experiments and Modeling in Support of Generic Salt Repository Science

Energy Technology Data Exchange (ETDEWEB)

Bourret, Suzanne Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stauffer, Philip H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weaver, Douglas James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Caporuscio, Florie Andre [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Otto, Shawn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Boukhalfa, Hakim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jordan, Amy B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chu, Shaoping [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zyvoloski, George Anthony [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Johnson, Peter Jacob [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-01-19

Salt is an attractive material for the disposition of heat generating nuclear waste (HGNW) because of its self-sealing, viscoplastic, and reconsolidation properties (Hansen and Leigh, 2012). The rate at which salt consolidates and the properties of the consolidated salt depend on the composition of the salt, including its content in accessory minerals and moisture, and the temperature under which consolidation occurs. Physicochemical processes, such as mineral hydration/dehydration salt dissolution and precipitation play a significant role in defining the rate of salt structure changes. Understanding the behavior of these complex processes is paramount when considering safe design for disposal of heat-generating nuclear waste (HGNW) in salt formations, so experimentation and modeling is underway to characterize these processes. This report presents experiments and simulations in support of the DOE-NE Used Fuel Disposition Campaign (UFDC) for development of drift-scale, in-situ field testing of HGNW in salt formations.

Hydrogen Generation in Microbial Reverse-Electrodialysis Electrolysis Cells Using a Heat-Regenerated Salt Solution

KAUST Repository

Nam, Joo-Youn

2012-05-01

Hydrogen gas can be electrochemically produced in microbial reverse-electrodialysis electrolysis cells (MRECs) using current derived from organic matter and salinity-gradient energy such as river water and seawater solutions. Here, it is shown that ammonium bicarbonate salts, which can be regenerated using low-temperature waste heat, can also produce sufficient voltage for hydrogen gas generation in an MREC. The maximum hydrogen production rate was 1.6 m3 H2/m3·d, with a hydrogen yield of 3.4 mol H2/mol acetate at a salinity ratio of infinite. Energy recovery was 10% based on total energy applied with an energy efficiency of 22% based on the consumed energy in the reactor. The cathode overpotential was dependent on the catholyte (sodium bicarbonate) concentration, but not the salinity ratio, indicating high catholyte conductivity was essential for maximizing hydrogen production rates. The direction of the HC and LC flows (co- or counter-current) did not affect performance in terms of hydrogen gas volume, production rates, or stack voltages. These results show that the MREC can be successfully operated using ammonium bicarbonate salts that can be regenerated using conventional distillation technologies and waste heat making the MREC a useful method for hydrogen gas production from wastes. © 2012 American Chemical Society.

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

2017-11-01

This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

Risk analyses for disposing nonhazardous oil field wastes in salt caverns

Energy Technology Data Exchange (ETDEWEB)

Tomasko, D.; Elcock, D.; Veil, J.; Caudle, D.

1997-12-01

Salt caverns have been used for several decades to store various hydrocarbon products. In the past few years, four facilities in the US have been permitted to dispose nonhazardous oil field wastes in salt caverns. Several other disposal caverns have been permitted in Canada and Europe. This report evaluates the possibility that adverse human health effects could result from exposure to contaminants released from the caverns in domal salt formations used for nonhazardous oil field waste disposal. The evaluation assumes normal operations but considers the possibility of leaks in cavern seals and cavern walls during the post-closure phase of operation. In this assessment, several steps were followed to identify possible human health risks. At the broadest level, these steps include identifying a reasonable set of contaminants of possible concern, identifying how humans could be exposed to these contaminants, assessing the toxicities of these contaminants, estimating their intakes, and characterizing their associated human health risks. The contaminants of concern for the assessment are benzene, cadmium, arsenic, and chromium. These were selected as being components of oil field waste and having a likelihood to remain in solution for a long enough time to reach a human receptor.

Training waste generators: The first responder in proper waste management

International Nuclear Information System (INIS)

Jones, E.

1989-01-01

Dealing with waste effectively requires a ''cradle to grave'' approach to waste management. The first step in that chain of custody is the waste generator. The waste generator plays the key role in the correct identification, packaging, and disposal of waste. The Technical Resources and Training Section at the Oak Ridge National Laboratory (ORNL) has developed several short training programs for waste generators. This training presents a consistent approach to proper handling of waste within the ORNL waste management system. This training has been developed for generators of solid low-level radioactive waste, hazardous and mixed waste, and transuranic waste. In addition to the above, a Waste Minimization training program has been developed for use by all organizations at ORNL who generate any type of hazardous waste. These training programs represent a combined effort of the training staff and the technical staff to assure that all ORNL staff accept their responsibility for handling all types of radioactive and hazardous wastes correctly from its generation to its disposal. 4 refs

Hydrometallurgical treatment of plutonium. Bearing salt baths waste

International Nuclear Information System (INIS)

Bros, P.; Gozlan, J.P.; Lecomte, M.; Bourges, J.

1993-01-01

The salt flux issuing from the electrorefining of plutonium metal alloy in salt baths (KCI + NaCI) poses a difficult problem of the back-end alpha waste management. An alternative to the salt process promoted by Los Alamos Laboratory is to develop a hydrometallurgical treatment. A new process based on the electrochemistry technique in aqueous solution has been defined and tested successfully in the CEA. The diagram of the process exhibits two principal steps: in the head-end, a dissolution in HNO 3 medium accompanied with an electrolytic dechlorination leading to a quantitative elimination of chloride as CI 2 gas followed by its trapping one soda lime cartridge, a complete oxidative dissolution of the refractory Pu residues by electrogenerated Ag(II), in the back-end: the Pu and Am recoveries by chromatographic extractions. (authors). 10 figs., 9 refs

Fundamental study on the salt distillation from the mixtures of rare earth precipitates and LiCl-KCl eutectic salt

International Nuclear Information System (INIS)

Yang, H. C.; Eun, H. C.; Cho, Y. Z.; Lee, H. S.; Kim, I. T.

2008-01-01

An electrorefining process of spent nuclear fuel generates waste salt containing some radioactive metal chlorides. The most effective method to reduce salt waste volume is to separate radioactive metals from non-radioactive salts. A promising approach is to change radioactive metal chlorides into salt-insoluble oxides by an oxygen sparging. Following this, salt distillation process is available to effectively separate the precipitated particulate metal oxides from salt. This study investigated the distillation rates of LiCl-KCl eutectic salt under different vacuums at elevated temperatures. The first part study investigated distillation rates of eutectic salt under different vacuums at high temperatures by using thermo-gravimetric furnace system. In the second part, we tested the removal of eutectic salt from the RE precipitates by using the laboratory vacuum distillation furnace system. Investigated variables were the temperature of mixture, the degree of vacuum and the time

Molten salt combustion of radioactive wastes

International Nuclear Information System (INIS)

Grantham, L.F.; McKenzie, D.E.; Richards, W.L.; Oldenkamp, R.D.

1976-01-01

The Atomics International Molten Salt Combustion Process reduces the weight and volume of combustible β-γ contaminated transuranic waste by utilizing air in a molten salt medium to combust organic materials, to trap particulates, and to react chemically with any acidic gases produced during combustion. Typically, incomplete combustion products such as hydrocarbons and carbon monoxide are below detection limits (i.e., 3 ) is directly related to the sodium chloride vapor pressure of the melt; >80% of the particulate is sodium chloride. Essentially all metal oxides (combustion ash) are retained in the melt, e.g., >99.9% of the plutonium, >99.6% of the europium, and >99.9% of the ruthenium are retained in the melt. Both bench-scale radioactive and pilot scale (50 kg/hr) nonradioactive combustion tests have been completed with essentially the same results. Design of three combustors for industrial applications are underway

Possible salt mine sites for radioactive waste disposal in the northeastern states

Energy Technology Data Exchange (ETDEWEB)

Landes, K.K.

1972-06-30

The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further.

Possible salt mine sites for radioactive waste disposal in the northeastern states

International Nuclear Information System (INIS)

Landes, K.K.

1972-01-01

The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further

Disposal alternatives and recommendations for waste salt management for repository excavation in the Palo Duro Basin, Texas

International Nuclear Information System (INIS)

1987-01-01

This report documents an evaluation of five alternatives for the disposal of waste salt that would be generated by the construction of a repository for radioactive waste in underground salt deposits at either of two sites in the Palo Duro Basin, Texas. The alternatives include commercial disposal, offsite deep-well injection, disposal in abandoned mines, ocean disposal, and land surface disposal on or off the site. For each alternative a reference case was rated - positive, neutral, or negative - in terms of environmental and dependability factors developed specifically for Texas sites. The factors constituting the environmental checklist relate to water quality impact, water- and land-use conflicts, ecological compatibility, conformity with air quality standards, and aesthetic impact. Factors on the dependability check-list relate to public acceptance, the adequacy of site characterization, permit and licensing requirements, technological requirements, and operational availability. A comparison of the ratings yielded the following viable alternatives, in order of preference: (1) land surface disposal, specifically disposal on tailings piles associated with abandoned potash mines; (2) disposal in abandoned mines, specifically potash mines; and (3) commercial disposal. Approaches to the further study of these three salt management techniques are recommended

Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 8. Repository preconceptual design studies: salt

International Nuclear Information System (INIS)

1978-04-01

This volume, Volume 8 ''Repository Preconceptual Design Studies: Salt,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area, and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/9, ''Drawings for Repository Preconceptual Design Studies: Salt.''

Glovebox design requirements for molten salt oxidation processing of transuranic waste

International Nuclear Information System (INIS)

Ramsey, K.B.; Acosta, S.V.; Wernly, K.D.

1998-01-01

This paper presents an overview of potential technologies for stabilization of 238 Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from 238 Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

Energy Technology Data Exchange (ETDEWEB)

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

2017-11-01

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

Waste generator services implementation plan

Energy Technology Data Exchange (ETDEWEB)

Mousseau, J.; Magleby, M.; Litus, M.

1998-04-01

Recurring waste management noncompliance problems have spurred a fundamental site-wide process revision to characterize and disposition wastes at the Idaho National Engineering and Environmental Laboratory. The reengineered method, termed Waste Generator Services, will streamline the waste acceptance process and provide waste generators comprehensive waste management services through a single, accountable organization to manage and disposition wastes in a timely, cost-effective, and compliant manner. This report outlines the strategy for implementing Waste Generator Services across the INEEL. It documents the culmination of efforts worked by the LMITCO Environmental Management Compliance Reengineering project team since October 1997. These efforts have included defining problems associated with the INEEL waste management process; identifying commercial best management practices; completing a review of DOE Complex-wide waste management training requirements; and involving others through an Integrated Process Team approach to provide recommendations on process flow, funding/charging mechanisms, and WGS organization. The report defines the work that will be performed by Waste Generator Services, the organization and resources, the waste acceptance process flow, the funding approach, methods for measuring performance, and the implementation schedule and approach. Field deployment will occur first at the Idaho Chemical Processing Plant in June 1998. Beginning in Fiscal Year 1999, Waste Generator Services will be deployed at the other major INEEL facilities in a phased approach, with implementation completed by March 1999.

Waste generator services implementation plan

International Nuclear Information System (INIS)

Mousseau, J.; Magleby, M.; Litus, M.

1998-04-01

Recurring waste management noncompliance problems have spurred a fundamental site-wide process revision to characterize and disposition wastes at the Idaho National Engineering and Environmental Laboratory. The reengineered method, termed Waste Generator Services, will streamline the waste acceptance process and provide waste generators comprehensive waste management services through a single, accountable organization to manage and disposition wastes in a timely, cost-effective, and compliant manner. This report outlines the strategy for implementing Waste Generator Services across the INEEL. It documents the culmination of efforts worked by the LMITCO Environmental Management Compliance Reengineering project team since October 1997. These efforts have included defining problems associated with the INEEL waste management process; identifying commercial best management practices; completing a review of DOE Complex-wide waste management training requirements; and involving others through an Integrated Process Team approach to provide recommendations on process flow, funding/charging mechanisms, and WGS organization. The report defines the work that will be performed by Waste Generator Services, the organization and resources, the waste acceptance process flow, the funding approach, methods for measuring performance, and the implementation schedule and approach. Field deployment will occur first at the Idaho Chemical Processing Plant in June 1998. Beginning in Fiscal Year 1999, Waste Generator Services will be deployed at the other major INEEL facilities in a phased approach, with implementation completed by March 1999

Applicability of molten salt oxidation to the destruction of actinide-contaminated wastes

International Nuclear Information System (INIS)

West, M.H.; Garcia, E.; Griego, W.J.; Court, D.B.; Rodriguez, L.

1992-01-01

A 1989 ban on incineration in the state of New Mexico caused cessation of actinide-contaminated cheesecloth, paper, and wood incineration within the Plutonium Facility (TA-55) at Los Alamos National Laboratory. Subsequently, plastic wipes were substituted for cheesecloth in the cleaning of glovebox interiors. However, waste minimization is not achieved by these measures since the wipes are discarded as Waste Isolation Pilot Plant certifiable wastes. After the ban was instituted, thermal decomposition of cheesecloth under argon at elevated temperature was examined and found satisfactory although scale of operation and speed were inferior to incineration. In 1991, the ban on incineration was lifted in New Mexico but Alamos has not chosen to pursue renewal of incineration at the Plutonium Facility. This paper reports that Los Alamos is looking from alternatives to incineration and thermal decomposition which are compatible with molten salt processing technology, historically a strength in actinide research at the Laboratory. Also, the technology must significantly reduce the volume of the waste upon treatment, i.e. waste minimization. Molten salt oxidation (MSO) has the promise of such a technology

Glovebox design requirements for molten salt oxidation processing of transuranic waste

Energy Technology Data Exchange (ETDEWEB)

Ramsey, K.B.; Acosta, S.V. [Los Alamos National Lab., NM (United States); Wernly, K.D. [Molten Salt Oxidation Corp., Bensalem, PA (United States)

1998-12-31

This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented.

Recovery of plutonium and americium from chloride salt wastes by solvent extraction

International Nuclear Information System (INIS)

Reichley-Yinger, L.; Vandegrift, G.F.

1987-01-01

Plutonium and americium can be recovered from aqueous waste solutions containing a mixture of HCl and chloride salt wastes by the coupling of two solvent extraction systems: tributyl phosphate (TBP) in tetrachloroethylene (TCE) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in TCE. In the flowsheet developed, the salt wastes are dissolved in HCl, the Pu(III) is oxidized to the IV state with NaClO 2 and recovered in the TBP-TCE cycle, and the Am is then removed from the resultant raffinate by the CMPO-TCE cycle. The consequences of the feed solution composition and extraction behavior of these species on the process flowsheet design, the Pu-product purity, and the decontamination of the aqueous raffinate from transuranic elements are discussed. 16 refs., 6 figs

MSO spent salt clean-up recovery process; TOPICAL

International Nuclear Information System (INIS)

Adamson, M G; Brummond, W A; Hipple, D L; Hsu, P C; Summers, L J; Von Holtz, E H; Wang, F T

1997-01-01

An effective process has been developed to separate metals, mineral residues, and radionuclides from spent salt, a secondary waste generated by Molten Salt Oxidation (MSO). This process includes salt dissolution, pH adjustment, chemical reduction and/or sulfiding, filtration, ion exchange, and drying. The process uses dithionite to reduce soluble chromate and/or sulfiding agent to suppress solubilities of metal compounds in water. This process is capable of reducing the secondary waste to less than 5% of its original weight. It is a low temperature, aqueous process and has been demonstrated in the laboratory[1

Modeling internal deformation of salt structures targeted for radioactive waste disposal

International Nuclear Information System (INIS)

Chemia, Zurab

2008-01-01

This thesis uses results of systematic numerical models to argue that externally inactive salt structures, which are potential targets for radioactive waste disposal, might be internally active due to the presence of dense layers or blocks within a salt layer. The three papers that support this thesis use the Gorleben salt diapir (NW Germany), which was targeted as a future final repository for high-grade radioactive waste, as a general guideline. The first two papers present systematic studies of the parameters that control the development of a salt diapir and how it entrains a dense anhydrite layer. Results from these numerical models show that the entrainment of a dense anhydrite layer within a salt diapir depends on four parameters: sedimentation rate, viscosity of salt, perturbation width and the stratigraphic location of the dense layer. The combined effect of these four parameters, which has a direct impact on the rate of salt supply (volume/area of the salt that is supplied to the diapir with time), shape a diapir and the mode of entrainment. Salt diapirs down-built with sedimentary units of high viscosity can potentially grow with an embedded anhydrite layer and deplete their source layer (salt supply ceases). However, when salt supply decreases dramatically or ceases entirely, the entrained anhydrite layer/segments start to sink within the diapir. In inactive diapirs, sinking of the entrained anhydrite layer is inevitable and strongly depends on the rheology of the salt, which is in direct contact with the anhydrite layer. During the post-depositional stage, if the effective viscosity of salt falls below the threshold value of around 10 18 -10 19 Pa s, the mobility of anhydrite blocks might influence any repository within the diapir. However, the internal deformation of the salt diapir by the descending blocks decreases with increase in effective viscosity of salt. The results presented in this thesis suggest that it is highly likely that salt structures

Definition of the waste package environment for a repository located in salt

International Nuclear Information System (INIS)

Clark, D.E.; Bradley, D.J.

1983-01-01

The expected environmental conditions for emplaced waste packages in a salt repository are simulated in the materials testing program to evaluate performance. Synthetic brines, based on the analyses of actual brines (both intrusion and inclusion), are used for corrosion and leach testing. Elevated temperatures (to 150 0 C) and radiation fields of up to 10 3 rad/h are employed as conservative conditions to bracket expected performance and provide data for worst case scenarios. Obtaining a precise definition of the waste package environment in a salt repository and its change with time is closely tied to detailed site characterization of the candidate salt repository horizon. It is expected that field testing can augment some of the materials testing currently under way and can provide increased confidence in the predicted site-specific near-field conditions. 17 references, 5 figures, 1 table

Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

International Nuclear Information System (INIS)

Bell, J.T.; Haas, P.A.; Rudolph, J.C.

1995-01-01

The Oak Ridge National Laboratory (ORNL) is participating in a program to apply a molten salt oxidation (MSO) process to treatment of mixed (radioactive and RCRA) wastes. The salt residues from the MSO treatment will require further separations or other processing to prepare them for final disposal. A bench-scale MSO apparatus is being installed at ORNL and will be operated on real Oak Ridge wastes. The treatment concepts to be tested and demonstrated on the salt residues from real wastes are described

Process to separate transuranic elements from nuclear waste

Science.gov (United States)

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1989-03-21

A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

Process to separate transuranic elements from nuclear waste

International Nuclear Information System (INIS)

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1989-01-01

A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs

Analyses of SRS waste glass buried in granite in Sweden and salt in the United States

International Nuclear Information System (INIS)

Williams, J.P.; Wicks, G.G.; Clark, D.E.; Lodding, A.R.

1991-01-01

Simulated Savannah River Site (SRS) waste glass forms have been buried in the granite geology of the Stirpa mine in Sweden for two years. Analyses of glass surfaces provided a measure of the performance of the waste glasses as a function of time. Similar SRS waste glass compositions have also been buried in salt at the WIPP facility in Carlsbad, New Mexico for a similar time period. Analyses of the SRS waste glasses buried in-situ in granite will be presented and compared to the performance of these same compositions buried in salt at WIPP

Forecasting the space-time stability of radioactive waste isolation in salt formations

International Nuclear Information System (INIS)

Anderson, E.B.; Karelin, A.I.; Krivokhatsiy, A.S.; Savonenkov, V.G.

1992-01-01

The possibilities to use salt formations for radioactive waste isolation are realized by creating shaft-type underground repositories in these rocks in Germany and the USA. The burial safety of low- and intermediate-level wastes for several hundred years have been substantiated for the sites chosen. Specialists of different countries presented positive properties of rock salt as a medium for isolation of radionuclides. A rich experience in building subsurface structures for different purposes in salts is accumulated in our country. Detailed investigations of salt formation have shown that far from all the saliferous areas and structures may be used for constructing burial sites. One of the reasons for this limitation is a sharp difference of individual deposits by their compositions, structures, the character of deposition and the conditions of formation. The geological criteria of safety acquire special significance in connection with the necessity to isolate radionuclides having the half-loves more than 1000 years. The time intervals required for stable isolation make up millions of years and cover great cycles of the evolution of the Earth surface and biosphere

Release consequence analysis for a hypothetical geologic radioactive waste repository in salt

International Nuclear Information System (INIS)

1979-08-01

One subtask conducted under the INFCE program is to evaluate and compare the health and safety impacts of different fuel cycles in which all radioactive wastes (except those from mining and milling) are placed in a geologic repository in salt. To achieve this objective, INFCE Working Group 7 examined the radiologic dose to humans from geologic repositories containing waste arisings as defined for seven reference fuel cycles. This report examines the release consequences for a generic waste repository in bedded salt. The top of the salt formation and the top of the repository are assumed to be 250 and 600 m, respectively, below the surface. The hydrogeologic structure above the salt consists of two aquifers and two aquitards. The aquifers connect to a river 6.2 km from the repository. The regional gradient to the river is 1 m/km in all aquifers. Hydrologic, transport, and dose models were used to model two release scenarios for each fuel cycle, one without a major disturbance and one in which a major geologic perturbation breached the repository immediately after it was sealed. The purpose of the modeling was to predict the rate of transport of radioactive contaminants from the repository through the geosphere to the biosphere, and to determine the potential dose to humans. Of the many radionuclides in the waste, only 129 I and 226 Ra arrived at the river in sufficient concentrations for a measurable dose calculation. Radionuclide concentrations in the ground water pose no threat to man because the ground water is a concentrated brine and it is diluted by a factor of 10 6 to 10 7 upon entering the river

Verifying generator waste certification: NTS waste characterization QA requirements

International Nuclear Information System (INIS)

Williams, R.E.; Brich, R.F.

1988-01-01

Waste management activities managed by the US Department of Energy (DOE) at the Nevada Test Site (NTS) include the disposal of low-level wastes (LLW) and mixed waste (MW), waste which is both radioactive and hazardous. A majority of the packaged LLW is received from offsite DOE generators. Interim status for receipt of MW at the NTS Area 5 Radioactive Waste Management Site (RWMS) was received from the state of Nevada in 1987. The RWMS Mixed Waste Management Facility (MWMF) is expected to be operational in 1988 for approved DOE MW generators. The Nevada Test Site Defense Waste Acceptance Criteria and Certification Requirements (NVO-185, Revision 5) delineates waste acceptance criteria for waste disposal at the NTS. Regulation of the hazardous component of mixed waste requires the implementation of US Environmental Protection Agency (EPA) requirements pursuant to the Resource Conservation and Recovery Act (RCRA). Waste generators must implement a waste certification program to provide assurance that the disposal site waste acceptance criteria are met. The DOE/Nevada Operations Office (NV) developed guidance for generator waste certification program plans. Periodic technical audits are conducted by DOE/NV to assess performance of the waste certification programs. The audit scope is patterned from the waste certification program plan guidance as it integrates and provides a common format for the applicable criteria. The criteria focus on items and activities critical to processing, characterizing, packaging, certifying, and shipping waste

Salt creep design consideration for underground nuclear waste storage

International Nuclear Information System (INIS)

Li, W.T.; Wu, C.L.; Antonas, N.J.

1983-01-01

This paper summarizes the creep consideration in the design of nuclear waste storage facilities in salt, describes the non-linear analysis method for evaluating the design adequacy, and presents computational results for the current storage design. The application of rock mechanics instrumentation to assure the appropriateness of the design is discussed. It also describes the design evolution of such a facility, starting from the conceptual design, through the preliminary design, to the detailed design stage. The empirical design method, laboratory tests and numerical analyses, and the underground in situ tests have been incorporated in the design process to assure the stability of the underground openings, retrievability of waste during the operation phase and encapsulation of waste after decommissioning

Salt splitting using ceramic membranes

Energy Technology Data Exchange (ETDEWEB)

Kurath, D.E. [Pacific Northwest National Lab., Richland, WA (United States)

1997-10-01

Many radioactive aqueous wastes in the DOE complex have high concentrations of sodium that can negatively affect waste treatment and disposal operations. Sodium can decrease the durability of waste forms such as glass and is the primary contributor to large disposal volumes. Waste treatment processes such as cesium ion exchange, sludge washing, and calcination are made less efficient and more expensive because of the high sodium concentrations. Pacific Northwest National Laboratory (PNNL) and Ceramatec Inc. (Salt Lake City UT) are developing an electrochemical salt splitting process based on inorganic ceramic sodium (Na), super-ionic conductor (NaSICON) membranes that shows promise for mitigating the impact of sodium. In this process, the waste is added to the anode compartment, and an electrical potential is applied to the cell. This drives sodium ions through the membrane, but the membrane rejects most other cations (e.g., Sr{sup +2}, Cs{sup +}). The charge balance in the anode compartment is maintained by generating H{sup +} from the electrolysis of water. The charge balance in the cathode is maintained by generating OH{sup {minus}}, either from the electrolysis of water or from oxygen and water using an oxygen cathode. The normal gaseous products of the electrolysis of water are oxygen at the anode and hydrogen at the cathode. Potentially flammable gas mixtures can be prevented by providing adequate volumes of a sweep gas, using an alternative reductant or destruction of the hydrogen as it is generated. As H{sup +} is generated in the anode compartment, the pH drops. The process may be operated with either an alkaline (pH>12) or an acidic anolyte (pH <1). The benefits of salt splitting using ceramic membranes are (1) waste volume reduction and reduced chemical procurement costs by recycling of NaOH; and (2) direct reduction of sodium in process streams, which enhances subsequent operations such as cesium ion exchange, calcination, and vitrification.

Reduced waste generation technical work plan

International Nuclear Information System (INIS)

1987-05-01

The United States Department of Energy has established policies for avoiding plutonium losses to the waste streams and minimizing the generation of wastes produced at its nuclear facilities. This policy is evidenced in DOE Order 5820.2, which states ''Technical and administrative controls shall be directed towards reducing the gross volume of TRU waste generated and the amount of radioactivity in such waste.'' To comply with the DOE directive, the Defense Transuranic Waste Program (DTWP) supports and provides funding for specific research and development tasks at the various DOE sites to reduce the generation of waste. This document has been prepared to give an overview of current and past Reduced Waste Generation task activities which are to be based on technical and cost/benefit factors. The document is updated annually, or as needed, to reflect the status of program direction. Reduced Waste Generation (RWG) tasks encompass a wide range of goals which are basically oriented toward (1) avoiding the generation of waste, (2) changing processes or operations to reduce waste, (3) converting TRU waste into LLW by sorting or decontamination, and (4) reducing volumes through operations such as incineration or compaction

Corrosion testing of selected packaging materials for disposal of high-level waste glass in rock salt formations

International Nuclear Information System (INIS)

Smailos, E.; Schwarzkopf, W.; Koester, R.; Fiehn, B.; Halm, G.

1990-05-01

In previous corrosion studies performed in salt brines, unalloyed steels, Ti 99.8-Pd and Hastelloy C4 have proved to be the most promising materials for long-term resistant packagings to be used in heat-generating waste (vitrified HLW, spent fuel) disposal in rock-salt formations. To characterise the corrosion behaviour of these materials in more detail, further in-depth laboratory-scale and in-situ corrosion studies have been performed in the present study. Besides the above-mentioned materials, also some in-situ investigations of the iron-base materials Ni-Resist D2 and D4, cast iron and Si-cast iron have been carried out in order to complete the results available to date. (orig.) [de

Performance Assessment of a Generic Repository in Bedded Salt for DOE-Managed Nuclear Waste

Science.gov (United States)

Stein, E. R.; Sevougian, S. D.; Hammond, G. E.; Frederick, J. M.; Mariner, P. E.

2016-12-01

A mined repository in salt is one of the concepts under consideration for disposal of DOE-managed defense-related spent nuclear fuel (SNF) and high level waste (HLW). Bedded salt is a favorable medium for disposal of nuclear waste due to its low permeability, high thermal conductivity, and ability to self-heal. Sandia's Generic Disposal System Analysis framework is used to assess the ability of a generic repository in bedded salt to isolate radionuclides from the biosphere. The performance assessment considers multiple waste types of varying thermal load and radionuclide inventory, the engineered barrier system comprising the waste packages, backfill, and emplacement drifts, and the natural barrier system formed by a bedded salt deposit and the overlying sedimentary sequence (including an aquifer). The model simulates disposal of nearly the entire inventory of DOE-managed, defense-related SNF (excluding Naval SNF) and HLW in a half-symmetry domain containing approximately 6 million grid cells. Grid refinement captures the detail of 25,200 individual waste packages in 180 disposal panels, associated access halls, and 4 shafts connecting the land surface to the repository. Equations describing coupled heat and fluid flow and reactive transport are solved numerically with PFLOTRAN, a massively parallel flow and transport code. Simulated processes include heat conduction and convection, waste package failure, waste form dissolution, radioactive decay and ingrowth, sorption, solubility limits, advection, dispersion, and diffusion. Simulations are run to 1 million years, and radionuclide concentrations are observed within an aquifer at a point approximately 4 kilometers downgradient of the repository. The software package DAKOTA is used to sample likely ranges of input parameters including waste form dissolution rates and properties of engineered and natural materials in order to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia

Possible salt mine and brined cavity sites for radioactive waste disposal in the northeastern southern peninsula of Michigan

International Nuclear Information System (INIS)

Landes, K.K.; Bourne, H.L.

1976-01-01

A reconnaissance report on the possibilities for disposal of radioactive waste covers Michigan only, and is more detailed than an earlier one involving the northeastern states. Revised ''ground rules'' for pinpointing both mine and dissolved salt cavern sites for waste disposal include environmental, geologic, and economic factors. The Michigan basin is a structural bowl of Paleozoic sediments resting on downwarped Precambrian rocks. The center of the bowl is in Clare and Gladwin Counties, a short distance north of the middle of the Southern Peninsula. The strata dip toward this central area, and some stratigraphic sequences, including especially the salt-containing Silurian section, increase considerably in thickness in that direction. Lesser amounts of salt are also present in the north central part of the Lower Peninsula. Michigan has been an oil and gas producing state since 1925 and widespread exploration has had two effects on the selection of waste disposal sites: (1) large areas are leased for oil and gas; and (2) the borehole concentrations, whether producing wells, dry holes, or industrial brine wells that penetrated the salt section, should be avoided. Two types of nuclear waste, low level and high level, can be stored in man-made openings in salt beds. The storage facilities are created by (1) the development of salt mines where the depths are less than 3000 ft, and (2) cavities produced by pumping water into a salt bed, and bringing brine back out. The high level waste disposal must be confined to mines of limited depth, but the low level wastes can be accommodated in brine cavities at any depth. Seven potential prospects have been investigated and are described in detail

Ionic Liquids for Utilization of Waste Heat from Distributed Power Generation Systems

Energy Technology Data Exchange (ETDEWEB)

Joan F. Brennecke; Mihir Sen; Edward J. Maginn; Samuel Paolucci; Mark A. Stadtherr; Peter T. Disser; Mike Zdyb

2009-01-11

The objective of this research project was the development of ionic liquids to capture and utilize waste heat from distributed power generation systems. Ionic Liquids (ILs) are organic salts that are liquid at room temperature and they have the potential to make fundamental and far-reaching changes in the way we use energy. In particular, the focus of this project was fundamental research on the potential use of IL/CO2 mixtures in absorption-refrigeration systems. Such systems can provide cooling by utilizing waste heat from various sources, including distributed power generation. The basic objectives of the research were to design and synthesize ILs appropriate for the task, to measure and model thermophysical properties and phase behavior of ILs and IL/CO2 mixtures, and to model the performance of IL/CO2 absorption-refrigeration systems.

Nuclear waste repository simulation experiments. Asse salt mine: Annual report 1984

International Nuclear Information System (INIS)

Rothfuchs, T.; Feddersen, H.K.; Schwarzianeck, P.; Staupendahl, G.; Coyle, A.J.; Kalia, H.; Eckert, J.

1985-01-01

This is the Second Annual Report (1984) which describes experiments simulating a nuclear waste repository at the 800 meter-level of the Asse Salt Mine in the Federal Republic of Germany. The report describes the Asse Salt Mine, the test equipment, and the pretest properties of the salt in the mine and in the vicinity of the test area. Also included are test data for the first sixteen months of operation on the following: brine migration rates, thermal mechanical behavior of the salt (including room closure, stress readings and thermal profiles) and borehole gas pressures. In addition to field data laboratory analyses of results are also included in this report. The duration of the experiment will be two years, ending in December 1985. (orig.)

Corrosion aspects of high-level waste disposal in salt domes

International Nuclear Information System (INIS)

Roerbo, K.

1979-12-01

In the ELSAM/ELKRAT waste management project it is planned that the high-level waste is glassified, encapsuled in canisters and finally deposited in a deep hole drilled in a salt dome. In the present report corrosion aspects of the canisters after deposition are discussed. The chemical environment will probably be a limited amount of brine coming from brine inclusions in the surrounding salt and moving up against the temperature gradient, the temperature at the canister surface being in the range of 100-150degC. The possible types of corrosion and the expected corrosion rates for a number of potential canister materials (mild steel, austenitic and ferritic stainless steels, Ni-base alloys, copper, titanium and a few combinations of materials) are discussed. Mild steel (possibly combined with an inner layer of copper or titanium) might possibly be an appropriate choice of material for the canister. (author)

Radiological consequences of a human intrusion in a nuclear waste repository in a salt formation

International Nuclear Information System (INIS)

Jacquier, P.; Raimbault, P.

1989-07-01

The assessment of the consequences of human intrusion scenarios for a repository is very important for salt formations, since this material has an undeniable economic interest. In this work, the scenario considers the solution mining of salt for human consumption: salt is extracted from a cavern; by leaching, this cavern enlarges and uncovers the waste, which falls down into the sump. It was assumed that the intrusion takes place either 500 years or 2500 years after the closing of the repository. High-level vitrified waste or alpha cemented waste were considered. This paper displays the assumptions made and, using a simplified modelling of the phenomena, the estimation of the radiological consequences due to ingestion of contamined sals. A sensitivity/uncertainty analysis is presented which emphasizes several fields where experimental studies have to be pursued or launched [fr

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.

On the time-dependent behavior of a cylindrical salt dome with a high-level waste repository

International Nuclear Information System (INIS)

Prij, J.

1988-01-01

In a salt dome with a repository for high-level radioactive and heat-generating waste, thermal stresses develop. These stresses can influence the isolation capability of the salt dome if these stresses can initiate cracks or introduce movements along existing closed flaws. The influence of the thermomechanical properties of the rock salt and the surrounding rocks on the thermal stresses and the surface rise is discussed. This discussion is based on a number of finite element creep analyses of a homogeneous cylindrical salt dome. The parameters, varied in the analyses, are constants in the thermomechanical constitutive behavior of salt and rocks, and furthermore the thermal loading has been varied. It is shown that variations in the creep properties, which result in differences in creep strain rate of a factor of 100, have only a very limited influence on the thermal stresses and the surface rise. Of more importance is the elastic stiffness of the materials. In all creep analyses the thermal stresses in the salt are compressive and the shear stresses remain below 2 MPa. The results are evaluated using an analytical treatment. Based on this evaluation, it is shown that the observed trends in the numerical results have a more general character and are not strictly limited to the geometry chosen. It is concluded that the thermal stresses in the salt formation are not strongly dependent on the creep properties of the rock salt

High-Level Waste Salt Disposition Systems Engineering Team Final Report, Volumes I, II, and III

International Nuclear Information System (INIS)

Piccolo, S.F.

1999-01-01

This report describes the process used and results obtained by the High Level Waste Salt Disposition Systems Engineering Team to select a primary and backup alternative salt disposition method for the Savannah River Site

Chemical stability of salt cake in the presence of organic materials

International Nuclear Information System (INIS)

Beitel, G.A.

1976-04-01

High-level waste stored as salt cake is principally NaNO 3 . Some organic material is known to have been added to the waste tanks. It has been suggested that some of this organic material may have become nitrated and transformed to a detonable state. Arguments are presented to discount the presence of nitrated organics in the waste tanks. Nitrated organics generated accidentally usually explode at the time of formation. Detonation tests show that salt cake and ''worst-case'' organic mixtures are not detonable. Organic mixtures with salt cake are compared with black powder, a related exothermic reactant. Black-powder mixtures of widely varying composition can and do burn explosively; ignition temperatures are 300-450 0 C. However, black-powder-type mixes cannot be ignited by radiation and are shock-insensitive. Temperatures generated by radionuclide decay in the salt are below 175 0 C and would be incapable of igniting any of these mixtures. The expected effect of radiation on organics in the waste tanks is a slow dehydrogenation and depolymerization along with a slight increase in sensitivity to oxidation. The greatest explosion hazard, if any exists, is a hydrogen--oxygen explosion from water radiolysis, but the hydrogen must first be generated and then trapped so that the concentration of hydrogen can rise above 4 vol percent. This is impossible in salt cake. Final confirmation of the safety against organic-related explosive reactions in the salt cake will be based upon analytical determinations of organic concentrations. 12 tables, 5 fig

Analysis of the geological stability of a hypothetical radioactive waste repository in a bedded salt formation

International Nuclear Information System (INIS)

Tierney, M.S.; Lusso, F.; Shaw, H.R.

1978-01-01

This document reports on the development of mathematical models used in preliminary studies of the long-term safety of radioactive wastes deeply buried in bedded salt formations. Two analytical approaches to estimating the geological stability of a waste repository in bedded salt are described: (a) use of probabilistic models to estimate the a priori likelihoods of release of radionuclides from the repository through certain idealized natural and anthropogenic causes, and (b) a numerical simulation of certain feedback effects of emplacement of waste materials upon ground-water access to the repository's host rocks. These models are applied to an idealized waste repository for the sake of illustration

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)

Waste Generation Overview, Course 23263

International Nuclear Information System (INIS)

Simpson, Lewis Edward

2016-01-01

This course, Waste Generation Overview Live (COURSE 23263), provides an overview of federal and state waste management regulations, as well as Los Alamos National Laboratory (LANL) policies and procedures for waste management operations. The course covers the activities involved in the cradle-to-grave waste management process and focuses on waste characterization, waste compatibility determinations and classification, and the storage requirements for temporary waste accumulation areas at LANL. When you have completed this course, you will be able to recognize federal, state, and LANL environmental requirements and their impact on waste operations; recognize the importance of the cradle-to-grave waste management process; identify the roles and responsibilities of key LANL waste management personnel (e.g., Waste Generator, Waste Management Coordinator, Waste Stream Profile approver, and Waste Certification Official); characterize a waste stream to determine whether it meets the definition of a hazardous waste, as well as characterize the use and minimum requirements for use of acceptable knowledge (AK) for waste characterization and waste compatibility documentation requirements; and identify the requirements for setting up and managing temporary waste accumulation areas.

Waste Generation Overview, Course 23263

Energy Technology Data Exchange (ETDEWEB)

Simpson, Lewis Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-11-28

This course, Waste Generation Overview Live (COURSE 23263), provides an overview of federal and state waste management regulations, as well as Los Alamos National Laboratory (LANL) policies and procedures for waste management operations. The course covers the activities involved in the cradle-to-grave waste management process and focuses on waste characterization, waste compatibility determinations and classification, and the storage requirements for temporary waste accumulation areas at LANL. When you have completed this course, you will be able to recognize federal, state, and LANL environmental requirements and their impact on waste operations; recognize the importance of the cradle-to-grave waste management process; identify the roles and responsibilities of key LANL waste management personnel (e.g., Waste Generator, Waste Management Coordinator, Waste Stream Profile approver, and Waste Certification Official); characterize a waste stream to determine whether it meets the definition of a hazardous waste, as well as characterize the use and minimum requirements for use of acceptable knowledge (AK) for waste characterization and waste compatibility documentation requirements; and identify the requirements for setting up and managing temporary waste accumulation areas.

Thermal properties of fly ash substituted slag cement waste forms for disposal of Savannah River Plant salt waste

International Nuclear Information System (INIS)

Roy, D.M.; Kaushal, S.; Licastro, P.H.; Langton, C.A.

1985-01-01

Waste processing at the Savannah River Plant will involve reconstitution of the salts (NaNO 3 , NaNO 2 , NaOH, etc.) into a concentrated solution (32 weight percent salts) followed by solidification in a cement-based waste form for burial. The stability and mechanical durability of such a 'saltstone monolith' will depend largely on the temperature reached due to heat of hydration and the thermal properties of the waste form. Fly ash has been used as an inexpensive constituent and to moderate the hydration and setting processes so as to avoid reaching prohibitively high temperatures which could cause thermal stresses. Both high-calcium and low-calcium fly ashes have been studied for this purpose. Other constituents of these mixes include granulated blast furnace slag and finely crushed limestone. Adiabatic temperature increase and thermal conductivity of these mixes have been studied and related x-ray diffraction and scanning electron microscopy studies carried out to understand the hydration process

Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

Science.gov (United States)

Hite, R.J.

1977-01-01

The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2, is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt anticlines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as 'marker beds.' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement.

Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

International Nuclear Information System (INIS)

Hite, R.J.

1977-01-01

The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2 , is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt antilcines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as ''marker beds.'' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement

Laboratory scale vitrification of low-level radioactive nitrate salts and soils from the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Shaw, P.; Anderson, B.

1993-07-01

INEL has radiologically contaminated nitrate salt and soil waste stored above and below ground in Pad A and the Acid Pit at the Radioactive Waste Management Complex. Pad A contain uranium and transuranic contaminated potassium and sodium nitrate salts generated from dewatered waste solutions at the Rocky Flats Plant. The Acid Pit was used to dispose of liquids containing waste mineral acids, uranium, nitrate, chlorinated solvents, and some mercury. Ex situ vitrification is a high temperature destruction of nitrates and organics and immobilizes hazardous and radioactive metals. Laboratory scale melting of actual radionuclides containing INEL Pad A nitrate salts and Acid Pit soils was performed. The salt/soil/additive ratios were varied to determine the range of glass compositions (resulted from melting different wastes); maximize mass and volume reduction, durability, and immobilization of hazardous and radioactive metals; and minimize viscosity and offgas generation for wastes prevalent at INEL and other DOE sites. Some mixtures were spiked with additional hazardous and radioactive metals. Representative glasses were leach tested and showed none. Samples spiked with transuranic showed low nuclide leaching. Wasteforms were two to three times bulk densities of the salt and soil. Thermally co-processing soils and salts is an effective remediation method for destroying nitrate salts while stabilizing the radiological and hazardous metals they contain. The measured durability of these low-level waste glasses approached those of high-level waste glasses. Lab scale vitrification of actual INEL contaminated salts and soils was performed at General Atomics Laboratory as part of the INEL Waste Technology Development and Environmental Restoration within the Buried Waste Integrated Demonstration Program

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

Equipment evaluation for low density polyethylene encapsulated nitrate salt waste at the Rocky Flats Plant

International Nuclear Information System (INIS)

Yamada, W.I.; Faucette, A.M.; Jantzen, R.C.; Logsdon, B.W.; Oldham, J.H.; Saiki, D.M.; Yudnich, R.J.

1993-01-01

Mixed wastes at the Rocky Flats Plant (RFP) are subject to regulation by the Resource Conservation and Recovery Act (RCRA). Polymer solidification is being developed as a final treatment technology for several of these mixed wastes, including nitrate salts. Encapsulation nitrate salts with low density polyethylene (LDPE) has been the preliminary focus of the RFP polymer solidification effort. Literature reviews, industry surveys, and lab-scale and pilot-scale tests have been conducted to evaluate several options for encapsulating nitrate salts with LDPE. Most of the effort has focused on identifying compatible drying and extrusion technologies. Other processing options, specifically meltration and non-heated compounding machines, were also investigated. The best approach appears to be pretreatment of the nitrate salt waste brine in either a vertical or horizontal thin film evaporator followed by compounding of the dried waste with LDPE in an intermeshing, co-rotating, twin-screw extruder. Additional pilot-scale tests planned for the fall of 1993 should further support this recommendation. Preliminary evaluation work indicates that meltration is not possible at atmospheric pressure with the LDPE (Chevron PE-1409) provided by RFP. However, meltration should be possible at atmospheric pressure using another LDPE formulation with altered physical and rheological properties: Lower molecular weight and lower viscosity (Epoline C-15). Contract modifications are now in process to allow a follow-on pilot scale demonstration. Questions regarding changed safety and physical properties of the resultant LDPE waste form due to use of the Epoline C-15 will be addressed. No additional work with non-heated mixer compounder machines is planned at this time

Solubility and speciation of actinides in salt solutions and migration experiments of intermediate level waste in salt formations

International Nuclear Information System (INIS)

1986-01-01

A comprehensive study into the solubility of the actinides americium and plutonium in concentrated salt solutions, the release of radionuclides from various forms of conditioned ILW and the migration behaviour of these nuclides through geological material specific to the Gorleben site in Lower Saxony is described. A detailed investigation into the characterization of four highly concentrated salt solutions in terms of their pH, Eh, inorganic carbon contents and their densities is given and a series of experiments investigating the solubility of standard americium(III) and plutonium(IV) hydroxides in these solutions is described. Transuranic mobility studies for solutions derived from the standard hydroxides through salt and sand have shown the presence of at least two types of species present of widely differing mobility; one migrating with approximately the same velocity as the solvent front and the other strongly retarded. Actinide mobility data are presented and discussed for leachates derived from the simulated ILW in cement and data are also presented for the migration of the fission products in leachates derived from real waste solidified in cement and bitumen. Relatively high plutonium mobilities were observed in the case of the former and in the case of the real waste leachates, cesium was found to be the least retarded. The sorption of ruthenium was found to be largely associated with the insoluble residues of the natural rock salt rather than the halite itself. (orig./RB)

Novel waste printed circuit board recycling process with molten salt.

Science.gov (United States)

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going.

Novel waste printed circuit board recycling process with molten salt

Science.gov (United States)

Riedewald, Frank; Sousa-Gallagher, Maria

2015-01-01

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

Energy Technology Data Exchange (ETDEWEB)

Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D. [U.S. Nuclear Regulatory Commission (United States)

2013-07-01

As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in

NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

International Nuclear Information System (INIS)

Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D.

2013-01-01

As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in coordination with South

Geological disposal of heat generating radioactive waste

International Nuclear Information System (INIS)

1986-03-01

A number of options for the disposal of vitrified heat-generating radioactive waste are being studied to ensure that safe methods are available when the time comes for disposal operations to commence. This study has considered the feasibility of three designs for containers which would isolate the waste from the environment for a minimum period of 500 to 1000 years. The study was sub-divided into the following major sections: manufacturing feasibility; stress analysis; integrity in accidents; cost benefit review. The candidate container designs were taken from the results of a previous study by Ove Arup and Partners (1985) and were developed as the study progressed. Their major features can be summarised as follows: (A) a thin-walled corrosion-resistant metal shell filled with lead or cement grout. (B) an unfilled thick-walled carbon steel shell. (C) an unfilled carbon steel shell planted externally with corrosion-resistant metal. Reference repository conditions in clay, granite and salt, reference disposal operations and metals corrosion data have been taken from various European Community radioactive waste management research and engineering projects. The study concludes that design Types A and B are feasible in manufacturing terms but design Type C is not. It is recommended that model containers should be produced to demonstrate the proposed methods of manufacture and that they should be tested to validate the analytical techniques used. (author)

Charging generators for waste management costs

International Nuclear Information System (INIS)

Berry, J.B.; Homan, F.J.

1988-01-01

Implementation of a plan to charge waste management costs to the facility that generates such waste requires a long-term commitment and consistent administration. The benefit is that generators are provided the incentive to optimize waste management practices if the charges are appropriately applied. This paper summarizes (1) a plan to charge waste generators, (2) the administrative structure of the plan, (3) a comparison between the rate structure and changes in waste disposal operations, and (4) issues that have surfaced as the plan is implemented. 2 refs., 1 fig., 1 tab

Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's report on Functional Design Criteria for a Repository for High-Level Radioactive Waste

International Nuclear Information System (INIS)

Hambley, D.F.; Russell, J.E.; Busch, J.S.; Harrison, W.; Edgar, D.E.; Tisue, M.W.

1984-08-01

This report summarizes Argonne's review of the Office of Nuclear Waste Isolation's (ONWI's) draft report entitled Functional Design Criteria for High-Level Nuclear Waste Repository in Salt, dated January 23, 1984. Recommendations are given for improving the ONWI draft report

Processes and parameters involved in modeling radionuclide transport from bedded salt repositories. Final report. Technical memorandum

International Nuclear Information System (INIS)

Evenson, D.E.; Prickett, T.A.; Showalter, P.A.

1979-07-01

The parameters necessary to model radionuclide transport in salt beds are identified and described. A proposed plan for disposal of the radioactive wastes generated by nuclear power plants is to store waste canisters in repository sites contained in stable salt formations approximately 600 meters below the ground surface. Among the principal radioactive wastes contained in these canisters will be radioactive isotopes of neptunium, americium, uranium, and plutonium along with many highly radioactive fission products. A concern with this form of waste disposal is the possibility of ground-water flow occurring in the salt beds and endangering water supplies and the public health. Specifically, the research investigated the processes involved in the movement of radioactive wastes from the repository site by groundwater flow. Since the radioactive waste canisters also generate heat, temperature is an important factor. Among the processes affecting movement of radioactive wastes from a repository site in a salt bed are thermal conduction, groundwater movement, ion exchange, radioactive decay, dissolution and precipitation of salt, dispersion and diffusion, adsorption, and thermomigration. In addition, structural changes in the salt beds as a result of temperature changes are important. Based upon the half-lives of the radioactive wastes, he period of concern is on the order of a million years. As a result, major geologic phenomena that could affect both the salt bed and groundwater flow in the salt beds was considered. These phenomena include items such as volcanism, faulting, erosion, glaciation, and the impact of meteorites. CDM reviewed all of the critical processes involved in regional groundwater movement of radioactive wastes and identified and described the parameters that must be included to mathematically model their behavior. In addition, CDM briefly reviewed available echniques to measure these parameters

Reconsolidation of salt as applied to permanent seals for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Hansen, F.D.; Callahan, G.D.; Van Sembeek, L.L.

1993-01-01

Reconsolidated salt is a fundamental component of the permanent seals for the Waste Isolation Pilot Plant. As regulations are currently understood and seal concepts envisioned, emplaced salt is the sole long-term seal component designed to prevent the shafts from becoming preferred pathways for rating gases or liquids. Studies under way in support of the sealing function of emplaced salt include laboratory testing of crushed salt small-scale in situ tests, constitutive modeling of crushed salt, calculations of the opening responses during operation and closure, and design practicalities including emplacement techniques. This paper briefly summarizes aspects of these efforts and key areas of future work

Removal of uranium from spent salt from the moltensalt oxidation process

International Nuclear Information System (INIS)

Summers, L.; Hsu, P.C.; Holtz, E.V.; Hipple, D.; Wang, F.; Adamson, M.

1997-03-01

Molten salt oxidation (MSO) is a thermal process that has the capability of destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials. In this process, combustible waste and air are introduced into the molten sodium carbonate salt. The organic constituents of the waste materials are oxidized to carbon dioxide and water, while most of the inorganic constituents, including toxic metals, minerals, and radioisotopes, are retained in the molten salt bath. As these impurities accumulate in the salt, the process efficiency drops and the salt must be replaced. An efficient process is needed to separate these toxic metals, minerals, and radioisotopes from the spent carbonate to avoid generating a large volume of secondary waste. Toxic metals such as cadmium, chromium, lead, and zinc etc. are removed by a method described elsewhere. This paper describes a separation strategy developed for radioisotope removal from the mixed spent salt, as well as experimental results, as part of the spent salt cleanup. As the MSO system operates, inorganic products resulting from the reaction of halides, sulfides, phosphates, metals and radionuclides with carbonate accumulate in the salt bath. These must be removed to prevent complete conversion of the sodium carbonate, which would result in eventual losses of destruction efficiency and acid scrubbing capability. There are two operational modes for salt removal: (1) during reactor operation a slip-stream of molten salt is continuously withdrawn with continuous replacement by carbonate, or (2) the spent salt melt is discharged completely and the reactor then refilled with carbonate in batch mode. Because many of the metals and/or radionuclides captured in the salt are hazardous and/or radioactive, spent salt removed from the reactor would create a large secondary waste stream without further treatment. A spent salt clean up/recovery system is necessary to segregate these materials and minimize the amount of

Implications of thermophysical properties in geoscientific investigations for the disposal of nuclear waste in a salt dome

International Nuclear Information System (INIS)

Kopietz, J.

1984-01-01

Examples from laboratory and in-situ experiments on the thermomechanical behavior of rock salt are used to discuss the implications of thermophysical properties for disposal of nuclear waste in a salt dome. The implications of thermophysical properties are also illustrated by a brief review of geothermal investigations made within the scope of geological and hydrogeological exploration of the Gorleben salt dome in northern Germany. High-resolution temperature measurements performed in shallow and deep boreholes drilled for the exploration of the Gorleben salt dome, together with thermal conductivity measurements on representative core samples from these boreholes, are contributing to a determination of groundwater flow in the covering layers of the salt dome and to the identification of zones of impurity (eg carnallitite layers) within the salt structure. Data from these experiments are used for setting up numerical models for heat propagation around a prospective waste repository in the Gorleben salt dome. Long-term creep experiments on samples of rock salt at up to 400 deg C are used to derive constitutive relations on the creep behavior of salt. In-situ heating experiments are being conducted in the Asse salt mine to determine the effect of a heat source on the integrity of the surrounding salt rock. (author)

Defense Waste Processing Facility (DWPF), Modular CSSX Unit (CSSX), and Waste Transfer Line System of Salt Processing Program (U)

International Nuclear Information System (INIS)

CHANG, ROBERT

2006-01-01

All of the waste streams from ARP, MCU, and SWPF processes will be sent to DWPF for vitrification. The impact these new waste streams will have on DWPF's ability to meet its canister production goal and its ability to support the Salt Processing Program (ARP, MCU, and SWPF) throughput needed to be evaluated. DWPF Engineering and Operations requested OBU Systems Engineering to evaluate DWPF operations and determine how the process could be optimized. The ultimate goal will be to evaluate all of the Liquid Radioactive Waste (LRW) System by developing process modules to cover all facilities/projects which are relevant to the LRW Program and to link the modules together to: (1) study the interfaces issues, (2) identify bottlenecks, and (3) determine the most cost effective way to eliminate them. The results from the evaluation can be used to assist DWPF in identifying improvement opportunities, to assist CBU in LRW strategic planning/tank space management, and to determine the project completion date for the Salt Processing Program

Permian salt dissolution, alkaline lake basins, and nuclear-waste storage, Southern High Plains, Texas and New Mexico

International Nuclear Information System (INIS)

Reeves, C.C. Jr.; Temple, J.M.

1986-01-01

Areas of Permian salt dissolution associated with 15 large alkaline lake basins on and adjacent to the Southern High Plains of west Texas and eastern New Mexico suggest formation of the basins by collapse of strata over the dissolution cavities. However, data from 6 other alkaline basins reveal no evidence of underlying salt dissolution. Thus, whether the basins were initiated by subsidence over the salt dissolution areas or whether the salt dissolution was caused by infiltration of overlying lake water is conjectural. However, the fact that the lacustrine fill in Mound Lake greatly exceeds the amount of salt dissolution and subsidence of overlying beds indicates that at least Mound Lake basin was antecedent to the salt dissolution. The association of topography, structure, and dissolution in areas well removed from zones of shallow burial emphasizes the susceptibility of Permian salt-bed dissolution throughout the west Texas-eastern New Mexico area. Such evidence, combined with previous studies documenting salt-bed dissolution in areas surrounding a proposed high-level nuclear-waste repository site in Deaf Smith County, Texas, leads to serious questions about the rationale of using salt beds for nuclear-waste storage

Salt Repository Project: Waste Package Program (WPP) modeling activiteis: FY 1984 annual report

International Nuclear Information System (INIS)

Kuhn, W.L.; Simonson, S.A.; Pulsipher, B.A.

1987-03-01

The Pacific Northwest Laboratory (PNL) is supporting the US Department of Energy's (DOE) Salt Repository Project (SRP) through its Waste Package Program (WPP). During FY 1984, the WPP continued its program of waste package component development and interactions testing and application of the resulting data base to develop predictive models describing waste package degradation and radionuclide release. Within the WPP, the Modeling Task (Task 04 during FY 1984) was conducted to interpret the tests in such a way that scientifically defensible models can be developed for use in qualification of the waste package

Treatment of radioactive waste salt by using synthetic silica-based phosphate composite for de-chlorination and solidification

Science.gov (United States)

Cho, In-Hak; Park, Hwan-Seo; Lee, Ki-Rak; Choi, Jung-Hun; Kim, In-Tae; Hur, Jin Mok; Lee, Young-Seak

2017-09-01

In the radioactive waste management, waste salts as metal chloride generated from a pyrochemical process to recover uranium and transuranic elements are one of problematic wastes due to their intrinsic properties such as high volatility and low compatibility with conventional glasses. This study reports a method to stabilize and solidify LiCl waste via de-chlorination using a synthetic composite, U-SAP (SiO2-Al2O3-B2O3-Fe2O3-P2O5) prepared by a sol-gel process. The composite was reacted with alkali metal elements to produce some metal aluminosilicates, aluminophosphates or orthophosphate as a crystalline or amorphous compound. Different from the original SAP (SiO2-Al2O3-P2O5), the reaction product of U-SAP could be successfully fabricated as a monolithic wasteform without a glassy binder at a proper reaction/consolidation condition. From the results of the FE-SEM, FT-IR and MAS-NMR analysis, it could be inferred that the Si-rich phase and P-rich phase as a glassy grains would be distributed in tens of nm scale, where alkali metal elements would be chemically interacted with Si-rich or P-rich region in the virgin U-SAP composite and its products was vitrified into a silicate or phosphate glass after a heat-treatment at 1150 °C. The PCT-A (Product Consistency Test, ASTM-1208) revealed that the mass loss of Cs and Sr in the U-SAP wasteform had a range of 10-3∼10-1 g/m2 and the leach-resistance of the U-SAP wasteform was comparable to other conventional wasteforms. From the U-SAP method, LiCl waste salt was effectively stabilized and solidified with high waste loading and good leach-resistance.

Stabilization/Solidification of radioactive molten salt waste by using xSiO{sub 2}-yAl{sub 2}O{sub 3}-zP{sub 2}O{sub 5} material

Energy Technology Data Exchange (ETDEWEB)

Hwan-Seo Park; In-Tae Kim; Yong-Zun Cho; Seong-Won Park; Eung-Ho Kim [Korea Atomic Energy Research Institute: 150 Deokjin-dong, Yuseong, Daejeon, 305-353 (Korea, Republic of)

2008-07-01

Molten salt waste generated from the electro metallurgical process to recover uranium and transuranic elements is considered as one of problematic wastes to be difficult to immobilize into a durable for final disposal. As an alternative, this study suggested a new method performed at molten state, where dechlorination was achieved with a new inorganic material containing SiO{sub 2}, Al{sub 2}O{sub 3} and P{sub 2}O{sub 5} (SAP). The SAP as a reactive material to molten salt was prepared by a conventional sol-gel process. The prepared SAPs were reacted with each metal chloride, LiCl, CsCl, SrCl{sub 2} and CeCl{sub 3} at 650 deg. C for 6 hours and also were reacted with simulated salt waste consisting of 90 wt% LiCl, 6.8 wt% CsCl and 3.2 wt% SrCl{sub 2} at different waste loading. All the reactions were carried out in oxidative atmosphere and metal chlorides were effectively converted into stable products under a reasonable reaction ratio.

Waste Generation in Denmark 1994-2005

DEFF Research Database (Denmark)

Brix, Louise Lykke; Bentzen, Jan Børsen

In recent years the amount of waste generated by Danish firms has increased significantly. In the present analysis we use the decomposition analysis, which has been widely used in the energy economics literature to explain the mechanisms influencing energy consumption and CO2 emissions. In this p......In recent years the amount of waste generated by Danish firms has increased significantly. In the present analysis we use the decomposition analysis, which has been widely used in the energy economics literature to explain the mechanisms influencing energy consumption and CO2 emissions....... In this paper the methodology is transferred to the topic of waste generation and is used to analyse why the amount of business waste is increasing. The empirical application is related to data for the volumes of waste generated in the Danish economy for the main sectors as well as the manufacturing sector...... covering the time span 1994-2005 has been included. By means of the Log-Mean Divisia Index Method (LMDI) an algebraic decomposition of the data for the waste amounts generated is performed. This methodology separates the increases in waste amounts into effects related to economic activity, industrial...

Estimation of construction and demolition waste using waste generation rates in Chennai, India.

Science.gov (United States)

Ram, V G; Kalidindi, Satyanarayana N

2017-06-01

A large amount of construction and demolition waste is being generated owing to rapid urbanisation in Indian cities. A reliable estimate of construction and demolition waste generation is essential to create awareness about this stream of solid waste among the government bodies in India. However, the required data to estimate construction and demolition waste generation in India are unavailable or not explicitly documented. This study proposed an approach to estimate construction and demolition waste generation using waste generation rates and demonstrated it by estimating construction and demolition waste generation in Chennai city. The demolition waste generation rates of primary materials were determined through regression analysis using waste generation data from 45 case studies. Materials, such as wood, electrical wires, doors, windows and reinforcement steel, were found to be salvaged and sold on the secondary market. Concrete and masonry debris were dumped in either landfills or unauthorised places. The total quantity of construction and demolition debris generated in Chennai city in 2013 was estimated to be 1.14 million tonnes. The proportion of masonry debris was found to be 76% of the total quantity of demolition debris. Construction and demolition debris forms about 36% of the total solid waste generated in Chennai city. A gross underestimation of construction and demolition waste generation in some earlier studies in India has also been shown. The methodology proposed could be utilised by government bodies, policymakers and researchers to generate reliable estimates of construction and demolition waste in other developing countries facing similar challenges of limited data availability.

Radioactive waste disposal in the Gorleben salt deposit

International Nuclear Information System (INIS)

Gizycki, P. von

1985-01-01

In the opinion of five experts, the protective function of the overlying rock as a barrier has turned out to be questionable after borings and measurements carried through at Gorleben. Moreover, the results have also raised doubts about the geological safety of the salt deposit as a barrier in the long run. The geological multibarrier concept must be discarded. Not only critics, but also 3 advocates from the field of official research on radioactive waste disposal state their opinion. (DG) [de

STORAGE AND RECOVERY OF SECONDARY WASTE COMING FROM MUNICIPAL WASTE INCINERATION PLANTS IN UNDERGROUND MINE

Directory of Open Access Journals (Sweden)

Waldemar Korzeniowski

2016-09-01

Full Text Available Regarding current and planned development of municipal waste incineration plants in Poland there is an important problem of the generated secondary waste management. The experience of West European countries in mining shows that waste can be stored successfully in the underground mines, but especially in salt mines. In Poland there is a possibility to set up the underground storage facility in the Salt Mine “Kłodawa”. The mine today is capable to locate over 3 million cubic meters and in the future it can increase significantly. Two techniques are proposed: 1 – storage of packaged waste, 2 – waste recovery as selfsolidifying paste with mining technology for rooms backfilling. Assuming the processing capacity of the storage facility as 100 000 Mg of waste per year, “Kłodawa” mine will be able to accept around 25 % of currently generated waste coming from the municipal waste incineration plants and the current volume of the storage space is sufficient for more than 20 years. Underground storage and waste recovery in mining techniques are beneficial for the economy and environment.

Cerebral salt wasting following traumatic brain injury

Directory of Open Access Journals (Sweden)

Peter Taylor

2017-04-01

Full Text Available Hyponatraemia is the most commonly encountered electrolyte disturbance in neurological high dependency and intensive care units. Cerebral salt wasting (CSW is the most elusive and challenging of the causes of hyponatraemia, and it is vital to distinguish it from the more familiar syndrome of inappropriate antidiuretic hormone (SIADH. Managing CSW requires correction of the intravascular volume depletion and hyponatraemia, as well as mitigation of on-going substantial sodium losses. Herein we describe a challenging case of CSW requiring large doses of hypertonic saline and the subsequent substantial benefit with the addition of fludrocortisone.

UK-Nuclear decommissioning authority and US Salt-stone waste management issues

International Nuclear Information System (INIS)

Lawless, William; Whitton, John

2007-01-01

Available in abstract form only. Full text of publication follows: We update two case studies of stakeholder issues in the UK and US. Earlier versions were reported at Waste Management 2006 and 2007 and at ICEM 2005. UK: The UK nuclear industry has begun to consult stakeholders more widely in recent years. Historically, methods of engagement within the industry have varied, however, recent discussions have generally been carried out with the explicit understanding that engagement with stakeholders will be 'dialogue based' and will 'inform' the final decision made by the decision maker. Engagement is currently being carried out at several levels within the industry; at the national level (via the Nuclear Decommissioning Authority's (NDA) National Stakeholder Group (NSG)); at a local site level (via Site Stakeholder Groups) and at a project level (usually via the Best Practicable Environmental Option process (BPEO)). This paper updates earlier results by the co-author with findings from a second questionnaire issued to the NSG in Phase 2 of the engagement process. An assessment is made regarding the development of stakeholder perceptions since Phase 1 towards the NDA process. US: The US case study reviews the resolution of issues on salt-stone by Department of Energy's (DOE) Savannah River Site (SRS) Citizens Advisory Board (CAB), in Aiken, SC. Recently, SRS-CAB encouraged DOE and South Carolina's regulatory Department of Health and Environmental Control (SC-DHEC) to resolve a conflict preventing SC-DHEC from releasing a draft permit to allow SRS to restart salt-stone operations. It arose with a letter sent from DOE blaming the Governor of South Carolina for delay in restarting salt processing. In reply, the Governor blamed DOE for failing to assure that Salt Waste Processing Facility (SWPF) would be built. SWPF is designed to remove most of the radioactivity from HLW prior to vitrification, the remaining fraction destined for salt-stone. (authors)

Brine Transport Experiments in Granular Salt

Energy Technology Data Exchange (ETDEWEB)

Jordan, Amy B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Boukhalfa, Hakim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Caporuscio, Florie Andre [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stauffer, Philip H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-06-06

To gain confidence in the predictive capability of numerical models, experimental validation must be performed to ensure that parameters and processes are correctly simulated. The laboratory investigations presented herein aim to address knowledge gaps for heat-generating nuclear waste (HGNW) disposal in bedded salt that remain after examination of prior field and laboratory test data. Primarily, we are interested in better constraining the thermal, hydrological, and physicochemical behavior of brine, water vapor, and salt when moist salt is heated. The target of this work is to use run-of-mine (RoM) salt; however during FY2015 progress was made using high-purity, granular sodium chloride.

Mass transfer in a salt repository

International Nuclear Information System (INIS)

Pigford, T.H.; Chambre, P.L.

1985-05-01

To meet regulatory requirements for radioactive waste in a salt repository it is necessary to predict the rates of corrosion of the waste container, the release rates of radionuclides from the waste package, and the cumulative release of radionuclides into the accessible environment. The mechanisms that may control these rates and an approach to predicting these rates from mass-transfer theory are described. This new mechanistic approach is suggested by three premises: (a) a brine inclusion originally in a salt crystal moves along grain boundaries after thermal-induced migration out of the crystal, (b) brine moves along a grain boundary under the influence of a pressure gradient, and (c) salt surrounding a heat-generating waste package will soon creep and consolidate as a monolithic medium surrounding and in contact with the waste package. After consolidation there may be very little migration of intergranular and intragranular brine to the waste package. The corrosion rate of the waste container may then be limited by the rate at which brine reaches the container and may be calculable from mass-transfer theory, and the rate at which dissolved radionuclides leave the waste package may be limited by molecular diffusion in intragranular brine and may be calculable from mass-transfer theory. If porous nonsalt interbeds intersect the waste-package borehole, the release rate of dissolved radionuclides to interbed brine may also be calculable from mass-transfer theory. The logic of these conclusions is described, as an aid in formulating the calculations that are to be made

The potential for using slags activated with near neutral salts as immobilisation matrices for nuclear wastes containing reactive metals

Science.gov (United States)

Bai, Y.; Collier, N. C.; Milestone, N. B.; Yang, C. H.

2011-06-01

The UK currently uses composite blends of Portland cement and other inorganic cementitious material such as blastfurnace slag and pulverised fuel ash to encapsulate or immobilise intermediate and low level radioactive wastes. Typically levels up 9:1 blast furnace slag:Portland cement or 4:1 pulverised fuel ash:Portland cement are used. Whilst these systems offer many advantages, their high pH causes corrosion of various metallic intermediate level radioactive wastes. To address this issue, lower pH/weakly alkaline cementitious systems have to be explored. While the blast furnace slag:Portland cement system is referred to as a composite cement system, the underlying reaction is actually an indirect activation of the slag hydration by the calcium hydroxide generated by the cement hydration, and by the alkali ions and gypsum present in the cement. However, the slag also can be activated directly with activators, creating a system known as alkali-activated slag. Whilst these activators used are usually strongly alkaline, weakly alkaline and near neutral salts can also be used. In this paper, the potential for using weakly alkaline and near neutral salts to activate slag in this manner is reviewed and discussed, with particular emphasis placed on the immobilisation of reactive metallic nuclear wastes.

Stabilization/Solidification of Radioactive LiCl-KCl Waste Salt by Using SiO2-Al2O3-P2O5(SAP) Inorganic Composite: Part 2. The Effect of SAP Composition on Stabilization/Solidification

International Nuclear Information System (INIS)

Ahn, Soo Na; Park, Hwan Seo; Cho, In Hak; Kim, In Tae; Cho, Yong Zun

2012-01-01

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of Fe 2 O 3 into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP(Fe=0.1). The experimental results indicated that the addition of Fe 2 O 3 increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing B 2 O 3 into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3 - 4 g of wasteform for final disposal. The final volume would be about 3 - 4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

The advantages of a salt/bentonite backfill for Waste Isolation Pilot Plant disposal rooms

International Nuclear Information System (INIS)

Butcher, B.M.; Novak, C.F.; Jercinovic, M.

1991-04-01

A 70/30 wt% salt/bentonite mixture is shown to be preferable to pure crushed salt as backfill for disposal rooms in the Waste Isolation Pilot Plant (WIPP). This report discusses several selection criteria used to arrive at this conclusion: the need for low permeability and porosity after closure, chemical stability with the surroundings, adequate strength to avoid shear erosion from human intrusion, ease of emplacement, and sorption potential for brine and radionuclides. Both salt and salt/bentonite are expected to consolidate to a final state of impermeability (i.e., ≤ 10 -18 m 2 ) adequate for satisfying federal nuclear regulations. Any advantage of the salt/bentonite mixture is dependent upon bentonite's potential for sorbing brine and radionuclides. Estimates suggest that bentonite's sorption potential for water in brine is much less than for pure water. While no credit is presently taken for brine sorption in salt/bentonite backfill, the possibility that some amount of inflowing brine would be chemically bound is considered likely. Bentonite may also sorb much of the plutonium, americium, and neptunium within the disposal room inventory. Sorption would be effective only if a major portion of the backfill is in contact with radioactive brine. Brine flow from the waste out through highly localized channels in the backfill would negate sorption effectiveness. Although the sorption potentials of bentonite for both brine and radionuclides are not ideal, they are distinctly beneficial. Furthermore, no detrimental aspects of adding bentonite to the salt as a backfill have been identified. These two observations are the major reasons for selecting salt/bentonite as a backfill within the WIPP. 39 refs., 16 figs., 6 tabs

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

Molten salt oxidation as a technique for decommissioning: selection of low melting point salt mixtures

International Nuclear Information System (INIS)

Lainetti, Paulo E.O.; Garcia, Vitor F.; Benvegnu, Guilherme

2013-01-01

During the 70 and 80 years, IPEN built several facilities in pilot scale, destined to the technological domain of the Nuclear Fuel Cycle. In the nineties, radical changes in the Brazilian nuclear policy determined the interruption of the activities and the shut-down of pilot plants. Nowadays, IPEN has been facing the problem of the dismantling and decommissioning of its Nuclear Fuel Cycle old facilities. The facility CELESTE-I of the IPEN is a laboratory where reprocessing studies were accomplished during the decade of 80 and in the beginning of the 90s. The last operations occurred in 92-93. The research activities generated radioactive wastes in the form of organic and aqueous solutions of different compositions and concentrations. For the treatment of these liquid wastes it was proposed a study of waste thermal decomposition based on the molten salt oxidation process.Decomposition tests of different organic wastes have been performed in laboratory equipment developed at IPEN, in the range of temperatures of 900 to 1020 deg C, demonstrating the complete oxidation of the compounds. The reduction of the process temperatures would be of crucial importance. Besides this, the selection of lower melting point salt mixtures would have an important impact in the reduction of equipment costs. Several experiments were performed to determine the most suitable salt mixtures, optimizing costs and melting temperatures as low as possible. This paper describes the main characteristics of the molten salt oxidation process, besides the selection of salt mixtures of binary and ternary compositions, respectively Na 2 CO 3 - NaOH and Na 2 CO 3 - K 2 CO 3 -Li 2 CO 3 . (author)

Numerical analysis of impurity separation from waste salt by investigating the change of concentration at the interface during zone refining process

Science.gov (United States)

Choi, Ho-Gil; Shim, Moonsoo; Lee, Jong-Hyeon; Yi, Kyung-Woo

2017-09-01

The waste salt treatment process is required for the reuse of purified salts, and for the disposal of the fission products contained in waste salt during pyroprocessing. As an alternative to existing fission product separation methods, the horizontal zone refining process is used in this study for the purification of waste salt. In order to evaluate the purification ability of the process, three-dimensional simulation is conducted, considering heat transfer, melt flow, and mass transfer. Impurity distributions and decontamination factors are calculated as a function of the heater traverse rate, by applying a subroutine and the equilibrium segregation coefficient derived from the effective segregation coefficients. For multipass cases, 1d solutions and the effective segregation coefficient obtained from three-dimensional simulation are used. In the present study, the topic is not dealing with crystal growth, but the numerical technique used is nearly the same since the zone refining technique was just introduced in the treatment of waste salt from nuclear power industry because of its merit of simplicity and refining ability. So this study can show a new application of single crystal growth techniques to other fields, by taking advantage of the zone refining multipass possibility. The final goal is to achieve the same high degree of decontamination in the waste salt as in zone freezing (or reverse Bridgman) method.

HAW project. Demonstrative disposal of high-level radioactive wastes in the Asse salt mine

International Nuclear Information System (INIS)

Rothfuchs, T.; Duijves, K.; Stippler, R.

1988-01-01

Since 1968 the GSF has been carrying out research and development programs for the final disposal of high-level radioactive waste (HAW) in salt formations. The heat producing waste has been simulated so far by means of electrical heaters and also cobalt-60-sources. In order to improve the final concept for HAW disposal in salt formations the complete technical system of an underground repository is to be tested in an one-to-one scale test facility. To satisfy the test objectives thirty high radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. The duration of testing will be approximately five years. For the handling of the radioactive canisters and their emplacement into the boreholes a system consisting of transportation casks, transportation vehicle, disposal machine, and borehole slider will be developed and tested. The actual scientific investigation program is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This program includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. The project is funded by the BMFT and the CEC and carrier out in close co-operation with the Netherlands Energy Research Foundation (ECN)

Waste segregation analysis for salt well pumping in the 200 W Area -- Task 3.4

International Nuclear Information System (INIS)

Reynolds, D.A.

1995-01-01

There is an estimated 7 million liters (1.9 million gallons) of potentially complexed waste that need to be pumped from single-shell tanks (SST) in the 200 West Area. This represents up to 40% of the salt well liquor that needs to be pumped in the 200 West Area. There are three double-shell (DST) tanks in the 241-SY tank farm in the 200 West Area. Tank 241-SY-101 is full and not usable. Tank 241-SY-102 has a transuranic (TRU) sludge in the bottom. Current rules prohibit mixing complexed waste with TRU waste. Tank 241-SY-103 has three major problems. First, 241-SY-103 is on the Flammable Watch list. Second, adding waste to tank 241-SY-103 has the potential for an episodic release of hydrogen gas. Third, 241-SY-103 will not hold all of the potentially complexed waste from the SSTs. This document looks at more details regarding the salt well pumping of the 200 West Area tank farm. Some options are considered but it is beyond the scope of this document to provide an in-depth study necessary to provide a defensible solution to the complexed waste problem

Waste Isolation Pilot Plant Salt Decontamination Testing

Energy Technology Data Exchange (ETDEWEB)

Rick Demmer; Stephen Reese

2014-09-01

On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

Polyethylene encapsulatin of nitrate salt wastes: Waste form stability, process scale-up, and economics

International Nuclear Information System (INIS)

Kalb, P.D.; Heiser, J.H. III; Colombo, P.

1991-07-01

A polyethylene encapsulation system for treatment of low-level radioactive, hazardous, and mixed wastes has been developed at Brookhaven National Laboratory. Polyethylene has several advantages compared with conventional solidification/stabilization materials such as hydraulic cements. Waste can be encapsulated with greater efficiency and with better waste form performance than is possible with hydraulic cement. The properties of polyethylene relevant to its long-term durability in storage and disposal environments are reviewed. Response to specific potential failure mechanisms including biodegradation, radiation, chemical attack, flammability, environmental stress cracking, and photodegradation are examined. These data are supported by results from extensive waste form performance testing including compressive yield strength, water immersion, thermal cycling, leachability of radioactive and hazardous species, irradiation, biodegradation, and flammability. The bench-scale process has been successfully tested for application with a number of specific ''problem'' waste streams. Quality assurance and performance testing of the resulting waste form confirmed scale-up feasibility. Use of this system at Rocky Flats Plant can result in over 70% fewer drums processed and shipped for disposal, compared with optimal cement formulations. Based on the current Rocky Flats production of nitrate salt per year, polyethylene encapsulation can yield an estimated annual savings between $1.5 million and $2.7 million, compared with conventional hydraulic cement systems. 72 refs., 23 figs., 16 tabs

Salt Separation from Uranium Deposits in Integrated Crucible

Energy Technology Data Exchange (ETDEWEB)

Kwon, S. W.; Park, K. M.; Chang, J. H.; Kim, J. G.; Park, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The solid cathode processing is necessary to separate the salt from the cathode since the uranium deposit in a solid cathode contains electrolyte salt. A physical separation process, such as distillation separation, is more attractive than a chemical or dissolution process because physical processes generate much less secondary process. Distillation process was employed for the cathode processsing due to the advantages of minimal generation of secondary waste, compact unit process, simple and low cost equipment. The basis for vacuum distillation separation is the difference in vapor pressures between salt and uranium. A solid cathode deposit is heated in a heating region and salt vaporizes, while non-volatile uranium remains behind. It is very important to increase the throughput of the salt separation system due to the high uranium content of spent nuclear fuel and high salt fraction of uranium dendrites. The evaporation rate of the LiCl-KCl eutectic salt in vacuum distiller is not so high to come up with the generation capacity of uranium dendrites in electro-refiner. Therefore, wide evaporation area or high distillation temperature is necessary for the successful salt separation. The adhered salt in the uranium deposits was removed successfully. The salt content in the deposits was below 0.1 wt% after the sequential operation of the liquid salt separation - salt distillation.

The HAW project. Demonstrative disposal of high-level radioactive wastes in the Asse salt mine

International Nuclear Information System (INIS)

Rothfuchs, T.; Duijves, K.

1988-04-01

Since 1968 the GSF has been carrying out research and development programs for the final disposal of high-level radioactive waste (HAW) in salt formations. The heat producing waste has been simulated so far by means of electrical heaters and also cobalt-60-sources. In order to improve the final concept for HAW disposal in salt formations the complete technical system of an underground repository is to be tested in a one-to-one scale test facility. To satisfy the test objectives thirty high radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. The duration of testing will be approximately five years. For the handling of the radioactive canisters and their emplacement into the boreholes a system consisting of transportation casks, transportation vehicle, disposal machine, and borehole slider will be developed and tested. The actual scientific investigation program is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This program includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./HP)

Separation of adhered salt from uranium deposits generated in electro-refiner

International Nuclear Information System (INIS)

Kwon, S.W.; Park, K.M.; Lee, H.S.; Kim, J.G.; Ahn, H.G.

2011-01-01

It is important to increase a throughput of the salt removal process from uranium deposits which is generated on the solid cathode of electro-refiner in pyroprocess. In this study, it was proposed to increase the throughput of the salt removal process by the separation of the liquid salt prior to the distillation of the LiCl-KCl eutectic salt from the uranium deposits. The feasibility of liquid salt separation was examined by salt separation experiments on a stainless steel sieve. It was found that the amount of salt to be distilled could be reduced by the liquid salt separation prior to the salt distillation. The residual salt remained in the deposits after the liquid salt separation was successfully removed further by the vacuum distillation. It was concluded that the combination of a liquid salt separation and a vacuum distillation is an effective route for the achievement of a high throughput performance in the salt separation process. (author)

Annual Report on Waste Generation and Waste Minimization Progress, 1991--1992

International Nuclear Information System (INIS)

1994-02-01

This report is DOE's first annual report on waste generation and waste minimization progress. Data presented in this report were collected from all DOE sites which met minimum threshold criteria established for this report. The fifty-seven site submittals contained herein represent data from over 100 reporting sites within 25 states. Radioactive, hazardous and sanitary waste quantities and the efforts to minimize these wastes are highlighted within the fifty-seven site submittals. In general, sites have made progress in moving beyond the planning phase of their waste minimization programs. This is evident by the overall 28 percent increase in the total amount of materials recycled from 1991 to 1992, as well as individual site initiatives. During 1991 and 1992, DOE generated a total of 279,000 cubic meters of radioactive waste and 243,000 metric tons of non-radioactive waste. These waste amounts include significant portions of process wastewater required to be reported to regulatory agencies in the state of Texas and the state of Tennessee. Specifically, the Pantex Plant in Texas treats an industrial wastewater that is considered by the Texas Water Commission to be a hazardous waste. In 1992, State regulated wastewater from the Pantex Plant represented 3,620 metric tons, 10 percent of the total hazardous waste generated by DOE. Similarly, mixed low-level wastewater from the TSCA Incinerator Facility at the Oak Ridge K-25 Site in Tennessee represented 55 percent of the total radioactive waste generated by DOE in 1992

Annual Report on Waste Generation and Waste Minimization Progress, 1991--1992

Energy Technology Data Exchange (ETDEWEB)

1994-02-01

This report is DOE`s first annual report on waste generation and waste minimization progress. Data presented in this report were collected from all DOE sites which met minimum threshold criteria established for this report. The fifty-seven site submittals contained herein represent data from over 100 reporting sites within 25 states. Radioactive, hazardous and sanitary waste quantities and the efforts to minimize these wastes are highlighted within the fifty-seven site submittals. In general, sites have made progress in moving beyond the planning phase of their waste minimization programs. This is evident by the overall 28 percent increase in the total amount of materials recycled from 1991 to 1992, as well as individual site initiatives. During 1991 and 1992, DOE generated a total of 279,000 cubic meters of radioactive waste and 243,000 metric tons of non-radioactive waste. These waste amounts include significant portions of process wastewater required to be reported to regulatory agencies in the state of Texas and the state of Tennessee. Specifically, the Pantex Plant in Texas treats an industrial wastewater that is considered by the Texas Water Commission to be a hazardous waste. In 1992, State regulated wastewater from the Pantex Plant represented 3,620 metric tons, 10 percent of the total hazardous waste generated by DOE. Similarly, mixed low-level wastewater from the TSCA Incinerator Facility at the Oak Ridge K-25 Site in Tennessee represented 55 percent of the total radioactive waste generated by DOE in 1992.

Laboratory creep and mechanical tests on salt data report (1975-1996): Waste Isolation Pilot Plant (WIPP) thermal/structural interactions program

Energy Technology Data Exchange (ETDEWEB)

Mellegard, K.D. [RE/SPEC Inc., Rapid City, SD (United States); Munson, D.E. [Sandia National Labs., Albuquerque, NM (United States)

1997-02-01

The Waste Isolation Pilot Plant (WIPP), a facility located in a bedded salt formation in Carlsbad, New Mexico, is being used by the U.S. Department of Energy to demonstrate the technology for safe handling and disposal of transuranic wastes produced by defense activities in the United States. In support of that demonstration, mechanical tests on salt were conducted in the laboratory to characterize material behavior at the stresses and temperatures expected for a nuclear waste repository. Many of those laboratory test programs have been carried out in the RE/SPEC Inc. rock mechanics laboratory in Rapid City, South Dakota; the first program being authorized in 1975 followed by additional testing programs that continue to the present. All of the WIPP laboratory data generated on salt at RE/SPEC Inc. over the last 20 years is presented in this data report. A variety of test procedures were used in performance of the work including quasi-static triaxial compression tests, constant stress (creep) tests, damage recovery tests, and multiaxial creep tests. The detailed data is presented in individual plots for each specimen tested. Typically, the controlled test conditions applied to each specimen are presented in a plot followed by additional plots of the measured specimen response. Extensive tables are included to summarize the tests that were performed. Both the tables and the plots contain cross-references to the technical reports where the data were originally reported. Also included are general descriptions of laboratory facilities, equipment, and procedures used to perform the work.

Permeability of natural rock salt from the Waste Isolation Pilot Plant (WIPP) during damage evolution and healing

International Nuclear Information System (INIS)

Pfeifle, T.W.; Hurtado, L.D.

1998-06-01

The US Department of Energy has developed the Waste Isolation Pilot Plant (WIPP) in the bedded salt of southeastern New Mexico to demonstrate the safe disposal of radioactive transuranic wastes. Four vertical shafts provide access to the underground workings located at a depth of about 660 meters. These shafts connect the underground facility to the surface and potentially provide communication between lithologic units, so they will be sealed to limit both the release of hazardous waste from and fluid flow into the repository. The seal design must consider the potential for fluid flow through a disturbed rock zone (DRZ) that develops in the salt near the shafts. The DRZ, which forms initially during excavation and then evolves with time, is expected to have higher permeability than the native salt. The closure of the shaft openings (i.e., through salt creep) will compress the seals, thereby inducing a compressive back-stress on the DRZ. This back-stress is expected to arrest the evolution of the DRZ, and with time will promote healing of damage. This paper presents laboratory data from tertiary creep and hydrostatic compression tests designed to characterize damage evolution and healing in WIPP salt. Healing is quantified in terms of permanent reduction in permeability, and the data are used to estimate healing times based on considerations of first-order kinetics

Municipal solid waste generation in Kathmandu, Nepal.

Science.gov (United States)

Dangi, Mohan B; Pretz, Christopher R; Urynowicz, Michael A; Gerow, Kenneth G; Reddy, J M

2011-01-01

Waste stream characteristics must be understood to tackle waste management problems in Kathmandu Metropolitan City (KMC), Nepal. Three-stage stratified cluster sampling was used to evaluate solid waste data collected from 336 households in KMC. This information was combined with data collected regarding waste from restaurants, hotels, schools and streets. The study found that 497.3 g capita(-1) day(-1) of solid waste was generated from households and 48.5, 113.3 and 26.1 kg facility(-1) day(-1) of waste was generated from restaurants, hotels and schools, respectively. Street litter measured 69.3 metric tons day(-1). The average municipal solid waste generation rate was 523.8 metric tons day(-1) or 0.66 kg capita(-1) day(-1) as compared to the 320 metric tons day(-1) reported by the city. The coefficient of correlation between the number of people and the amount of waste produced was 0.94. Key household waste constituents included 71% organic wastes, 12% plastics, 7.5% paper and paper products, 5% dirt and construction debris and 1% hazardous wastes. Although the waste composition varied depending on the source, the composition analysis of waste from restaurants, hotels, schools and streets showed a high percentage of organic wastes. These numbers suggest a greater potential for recovery of organic wastes via composting and there is an opportunity for recycling. Because there is no previous inquiry of this scale in reporting comprehensive municipal solid waste generation in Nepal, this study can be treated as a baseline for other Nepalese municipalities. Copyright © 2010 Elsevier Ltd. All rights reserved.

Charging generators for waste management costs

International Nuclear Information System (INIS)

Berry, J.B.; Homan, F.J.

1988-01-01

The Department of Energy (DOE) has recognized the need for waste management that incorporates improved waste-handling techniques and more stringent regulatory requirements to prevent future liabilities such as Superfund sites. DOE-Oak Ridge Operations (DOE-ORO) has recognized that an effective waste management program focuses on control at the source and that the burden for responsible waste management can be placed on generators by charging for waste management costs. The principle of including the waste management costs in the total cost of the product, even when the product is research and development, is being implemented at Oak Ridge National Laboratory (ORNL). This paper summarizes a plan to charge waste generators, the administrative structure of the plan, a comparison between the rate structure and changes in waste disposal operations, and issues that have surfaced as the plan is implemented

Temperature calculations on different configurations for disposal of high-level reprocessing waste in a salt dome model

International Nuclear Information System (INIS)

Hamstra, J.; Kevenaar, J.W.A.M.

1978-06-01

A medium size salt dome is considered as a structure in which a repository can be located for all radioactive wastes to be produced within the scope of a postulated nuclear power program. A dominating design factor for the lay-out of such a waste repository is the temperature distribution in the salt dome resulting from decay heat released from the buried solidified high-level reprocessing waste. Two numerical models are presented for the calculation of both global and local rock salt temperatures. The results of calculations performed with these models are demonstrated to be compatible. Rock salt temperatures related to several types of burial configurations, ranging from two layer configurations with various vertical distances between the layers via a three and a four layer repository to deep bore hole concepts varying from 100 to 600 m bore hole depth, can therefore be calculated with one rather simple unit cell model. The results of these calculations indicate that rock salt temperatures can be kept within acceptable limits to realize a repository using standard mining techniques. The temperatures at mine galery level prove to be a dominating factor in the selection of a repository configuration. More detailed calculations of these temperatures taking into account the loading sequence and the cooling capacity of the mine ventilation are recommended. Finally the apparent advantages of a deep bore hole concept emphasize the need for R and D work with respect to advanced drilling techniques in order to achieve deep dry disposal bore holes that can be realized from a burial mine in the salt dome. (Auth.)

Model for future waste generation

Energy Technology Data Exchange (ETDEWEB)

Sundqvist, Jan-Olov; Stenmarck, Aasa; Ekvall, Tomas

2010-06-15

The research presented in this report is part of the effort to estimate future Swedish waste quantities in the research programme Towards Sustainable Waste Management. More specifically, we estimate future waste coefficients that are designed to be fed into EMEC, which describes the Swedish economy in terms of 26 industrial sectors, a public sector, and households. Production in the model of industry and public sector requires input of labour, capital, energy, and other commodities. With waste-intensity coefficients added to each production parameter in each sector, EMEC can calculate the future waste quantities generated in different economic scenarios. To produce the waste-intensity coefficients, we make a survey of the current Swedish waste statistics. For each waste category from each sector we estimate whether the quantity depends primarily on the production in the sector, on the inputs of commodities, on the depreciation of capital goods, or on the size of the workforce in the sector. We calculate current waste-intensity coefficients by dividing the waste quantities by the parameter(s) to which they are assigned. We also present five different scenarios to describe how the waste intensity can develop until the year 2030. As far as possible and when deemed to be relevant, we have set the industrial waste generation to depend on the use of a commodity or an energy carrier. The quantity of spent vehicles and most equipment is set to depend on the depreciation of capital goods. Some wastes have been allocated to the staff, for example household waste from business. The quantities of wastes from households have a similar approach where every waste category is assigned to a combination of 26 different commodities

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.

Nuclear waste repository simulation experiments, Asse Salt Mine, Federal Republic of Germany. Annual report, 1983

International Nuclear Information System (INIS)

Rothfuchs, T.; Luebker, D.; Coyle, A.; Kalia, H.

1984-10-01

This is the First Annual report (1983) which describes experiments simulating a nuclear waste respository at the 800-meter level of the Asse Salt Mine in the Federal Republic of Germany. The report describes the test equipment, the Asse Salt Mine, the pretest properties of the salt in the test gallery, and the mine proper. Also included are test data for the first six months of operations on brine migration rates, room closure rates, extensometer readings, stress measurements, and thermal mechanical behavior of the salt. The duration of the experiments will be two years, ending in December 1985. 3 references, 34 figures, 13 tables

Summary Report of Comprehensive Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

Energy Technology Data Exchange (ETDEWEB)

Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-10-04

reactivity (D003). Follow-on testing was conducted to demonstrate the effectiveness of zeolite stabilization for ignitable WypAll and cheesecloth debris and additional nitrate salt solutions (those exhibiting the oxidizer characteristic) to demonstrate the effectiveness of the remedy. Follow-on testing also included testing of surrogate materials containing Waste Lock 770, which is present in four of the RNS containers, and potential items of debris such as plywood and Celotex material. Testing to evaluate the effectiveness of the remedy was performed using the specific remediation processes that are planned for use at the WCRRF. Finally, testing was also performed to evaluate the holding capacity of zeolite using a highly acidic surrogate solution and to characterize the composition of gases generated during mixing of zeolite with surrogate solutions. All these tests demonstrated the effectiveness of adding zeolite as the planned remedy.

The function of packing materials in a high-level nuclear waste repository and some candidate materials: Salt Repository Project

International Nuclear Information System (INIS)

Bunnell, L.R.; Shade, J.W.

1987-03-01

Packing materials should be included in waste package design for a high-level nuclear waste repository in salt. A packing material barrier would increase confidence in the waste package by alleviating possible shortcomings in the present design and prolonging confinement capabilities. Packing materials have been studied for uses in other geologic repositories; appropriately chosen, they would enhance the confinement capabilities of salt repository waste packages in several ways. Benefits of packing materials include retarding or chemically modifying brines to reduce corrosion of the waste package, providing good thermal conductivity between the waste package and host rock, retarding or absorbing radionuclides, and reducing the massiveness of the waste package. These benefits are available at low percentage of total repository cost, if the packing material is properly chosen and used. Several candidate materials are being considered, including oxides, hydroxides, silicates, cement-based mixtures, and clay mixtures. 18 refs

Cerebral salt wasting: a report of three cases

International Nuclear Information System (INIS)

Younas, H.; Sabir, O.; Tarif, N.

2015-01-01

Hyponatremia secondary to the Syndrome of Inappropriate Anti-Diuretic Hormone (SIADH) secretion is commonly observed in patients with various neurological disorders. Cerebral Salt Wasting (CSW) resulting in hyponatremia is also an infrequent occurrence in some patients with neurological disorders. Confusion in differentiating CSW from SIADH may arise since both results in similar electrolyte disturbances. Herein, we report three cases of CSW with intracranial afflictions. CSW was diagnosed on the basis of fractional excretion of urinary sodium and uric acid along with extremely low serum uric acid. Improvements in serum sodium levels after saline hydration and fludrocortisone administration further supported the diagnosis. (author)

Waste Generation Overview Refresher, Course 21464

Energy Technology Data Exchange (ETDEWEB)

Simpson, Lewis Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-12-13

This course, Waste Generation Overview Refresher (COURSE 21464), provides an overview of federal and state waste management regulations, as well as Los Alamos National Laboratory (LANL) policies and procedures for waste management operations. The course covers the activities involved in the cradle-to- grave waste management process and focuses on waste characterization, waste compatibility determinations and classification, and the storage requirements for temporary waste accumulation areas at LANL.

Mercury separation from aqueous wastes

International Nuclear Information System (INIS)

Taylor, P.A.; Klasson, K.T.; Corder, S.L.

1995-07-01

This project is providing an assessment of new sorbents for removing mercury from wastes at US Department of Energy sites. Four aqueous wastes were chosen for lab-scale testing; a high-salt, acidic waste currently stored at Idaho National Engineering Laboratory (INEL); a high-salt, alkaline waste stored at the Savannah River Site (SRS); a dilute lithium hydroxide solution stored at the Oak Ridge Y-12 Plant; and a low-salt, neutral groundwater generated at the Y-12 Plant. Eight adsorbents have been identified for testing, covering a wide range of cost and capability. Screening tests have been completed, which identified the most promising adsorbents for each waste stream. Batch isotherm tests have been completed using the most promising adsorbents, and column tests are in progress. Because of the wide range of waste compositions tested, no one adsorbent is effective in all of these waste streams. Based on loading capacity and compatibility with the waste solutions. the most effective adsorbents identified to date are SuperLig 618 for the INEL tank waste stimulant; Mersorb followed by lonac SR-3 for the SRS tank waste stimulant; Durasil 70 and Ionac SR-3) for the LIOH solution; and lonac SR-3 followed by lonac SR-4 and Mersorb for the Y-12 groundwater

Method of processing chloride waste

International Nuclear Information System (INIS)

Tokiwai, Moriyasu; Tsunashima, Mikiyasu; Horie, Masaaki; Koyama, Masafumi; Sudo, Minoru; Kitagawa, Masatoshi; Ogasawara, Tadashi.

1991-01-01

In a method of applying molten salt electrolysis to chloride wastes discharged from a electrolytic refining step of a dry reprocessing step for spent fuels, and removed with transuranium elements of long half-decaying time, metals capable of alloying with alkali and alkaline earth metals under melting by electrolysis are used as a cathode material, and an electrolytic temperature is made higher than the melting point of salts in a molten salt electrolysis bath, to recover Li, Ca and Na as alloys with the cathode material in a first electrolysis step. Then, the electrolytic temperature is made higher than the melting point of the chloride salts remained in the bath after the electrolysis step described above by using the cathode material, to recover Ba, Rb, Sr and Cs of nuclear fission products also as alloys with the cathode material in a second electrolysis step. Accordingly, the amount of wastes formed can be reduced, and the wastes contain no heat generating nuclear fission elements. (T.M.)

Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation

Energy Technology Data Exchange (ETDEWEB)

Morgan, Dane; Eapen, Jacob

2013-10-01

Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt

Annual report of waste generation and pollution prevention progress 1997

International Nuclear Information System (INIS)

1998-09-01

This sixth Annual Report presents and analyzes DOE Complex-wide waste generation and pollution prevention activities at 36 reporting sites from 1993 through 1997. In May 1996, the Secretary of Energy established a 50 percent Complex-Wide Waste Reduction Goal (relative to the 1993 baseline) for routine operations radioactive and hazardous waste generation, to be achieved by December 31, 1999. Excluding sanitary waste, routine operations waste generation increased three percent from 1996 to 1997, and decreased 61 percent overall from 1993 to 1997. DOE has achieved its Complex-Wide Waste Reduction Goals for routine operations based upon a comparison of 1997 waste generation to the 1993 baseline. However, it is important to note that increases in low-level radioactive and low-level mixed waste generation could reverse this achievement. From 1996 to 1997, low-level radioactive waste generation increased 10 percent, and low-level mixed waste generation increased slightly. It is critical that DOE sites continue to reduce routine operations waste generation for all waste types, to ensure that DOE's Complex-Wide Waste Reduction Goals are achieved by December 31, 1999

Radioactive waste isolation in salt: Peer review of the Golder Associates draft test plan for in situ testing in an exploratory shaft in salt

International Nuclear Information System (INIS)

Hambley, D.F.; Mraz, D.Z.; Unterberter, R.R.

1987-01-01

This report documents the peer review conducted by Argonne National Laboratory of a document entitled ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared for Battelle Memorial Institute's Office of Nuclear Waste Isolation by Golder Associates, Inc. In general, the peer review panelists found the test plan to be technically sound, although some deficiencies were identified. Recommendations for improving the test plan are presented in this review report. A microfiche copy of the following unpublished report is attached to the inside back cover of this report: ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared by Golder Associates, Inc., for Office of Nuclear Waste Isolation, Battelle Memorial Institute, Columbus, Ohio (March 1985)

Generic aspects of salt repositories

International Nuclear Information System (INIS)

Laughon, R.B.

1979-01-01

The history of geological disposal of radioactive wastes in salt is presented from 1957 when a panel of the National Academy of Sciences-National Research Council recommended burial in bedded salt deposits. Early work began in the Kansas, portion of the Permian Basin where simulated wastes were placed in an abandoned salt mine at Lyons, Kansas, in the late 1960's. This project was terminated when the potential effect of nearby solution mining activities could not be resolved. Evaluation of bedded salts resumed a few years later in the Permian Basin in southeastern New Mexico, and search for suitable sites in the 1970's resulted in the formation of the National Waste Terminal Storage Program in 1976. Evaluation of salt deposits in many regions of the United States has been virtually completed and has shown that deposits having the greatest potential for radioactive waste disposal are those of the largest depositional basins and salt domes of the Gulf Coast region

Mercury separation from mixed wastes. Annual report

International Nuclear Information System (INIS)

Taylor, P.A.; Klasson, K.T.; Corder, S.L.; Carlson, T.R.; McCandless, K.R.

1995-11-01

This is an assessment of new sorbents for removing Hg from wastes at US DOE sites. Four aqueous wastes were used for the laboratory tests: a simulant of a high-salt, acidic waste currently stored at INEL, a simulant of a high-salt, alkaline waste stored at Savannah River (SRS), a dilute LiOH solution stored at Y-12, and a low-salt, neutral groundwater generated at Y-12. Eight adsorbents covering a wide range of cost and capability were tested. Screening tests identified the most promising adsorbents, and column tests were performed using at least two adsorbents for each waste stream. No one adsorbent is effective in all of these waste streams. Based on loading capacity and compatibility, the most effect adsorbents to date are SuperLig 618 for the INEL tank waste simulant, Mersorb and Ionac SR-3 for the SRS tank waste simulant, Durasil 70 and Ionac SR-3 for the LiOH solution, and Ionac SR-3, followed by Ionac SR-4 and Mersorb, for the Y-12 groundwater

Aspects of the storage of radioactive waste

International Nuclear Information System (INIS)

Nienhuys, K.

1978-01-01

The expansion in the number of nuclear power stations in the netherlands is amongst other things, dependent on an acceptable policy for the storage of the waste from the stations. Consequently the idea has developed for storage in a salt-dome. The sub-committee on radioactive waste substances of the Interdepartmental Committee for Nuclear Energy has therefore given a mandate to initiate further research. For the risk analysis over the definitive storage of nuclear waste the sub-comittee produced a report in 1975, entitled 'Safety analysis for the underground storage of nuclear waste in salt-dome outcrops'. The analysis reveals a number of defective features. This makes especially clear that statements about the definitive storage of nuclear waste in salt domes can only be made with a great deal of uncertainty. There is no guarantee that the nuclear waste generated may be stowed away so that it will never return to the ionosphere. The speed whereby the nuclear waste may return would be dependent on a combination of events which cannot generally be calculated or assessed. The long term consequences of an irreversible radioactive contamination of the biosphere is not acceptable. There is insufficient proof that the storage of radioactive waste in salt domes is feasible. (G.C.)

Salton Sea Geothermal Field, California, as a near-field natural analog of a radioactive waste repository in salt

International Nuclear Information System (INIS)

Elders, W.A.; Cohen, L.H.

1983-11-01

Since high concentrations of radionuclides and high temperatures are not normally encountered in salt domes or beds, finding an exact geologic analog of expected near-field conditions in a mined nuclear waste repository in salt will be difficult. The Salton Sea Geothermal Field, however, provides an opportunity to investigate the migration and retardation of naturally occurring U, Th, Ra, Cs, Sr and other elements in hot brines which have been moving through clay-rich sedimentary rocks for up to 100,000 years. The more than thirty deep wells drilled in this field to produce steam for electrical generation penetrate sedimentary rocks containing concentrated brines where temperatures reach 365 0 C at only 2 km depth. The brines are primarily Na, K, Ca chlorides with up to 25% of total dissolved solids; they also contain high concentrations of metals such as Fe, Mn, Li, Zn, and Pb. This report describes the geology, geophysics and geochemistry of this system as a prelude to a study of the mobility of naturally occurring radionuclides and radionuclide analogs within it. The aim of this study is to provide data to assist in validating quantitative models of repository behavior and to use in designing and evaluating waste packages and engineered barriers. 128 references, 33 figures, 13 tables

Salton Sea Geothermal Field, California, as a near-field natural analog of a radioactive waste repository in salt

Energy Technology Data Exchange (ETDEWEB)

Elders, W.A.; Cohen, L.H.

1983-11-01

Since high concentrations of radionuclides and high temperatures are not normally encountered in salt domes or beds, finding an exact geologic analog of expected near-field conditions in a mined nuclear waste repository in salt will be difficult. The Salton Sea Geothermal Field, however, provides an opportunity to investigate the migration and retardation of naturally occurring U, Th, Ra, Cs, Sr and other elements in hot brines which have been moving through clay-rich sedimentary rocks for up to 100,000 years. The more than thirty deep wells drilled in this field to produce steam for electrical generation penetrate sedimentary rocks containing concentrated brines where temperatures reach 365/sup 0/C at only 2 km depth. The brines are primarily Na, K, Ca chlorides with up to 25% of total dissolved solids; they also contain high concentrations of metals such as Fe, Mn, Li, Zn, and Pb. This report describes the geology, geophysics and geochemistry of this system as a prelude to a study of the mobility of naturally occurring radionuclides and radionuclide analogs within it. The aim of this study is to provide data to assist in validating quantitative models of repository behavior and to use in designing and evaluating waste packages and engineered barriers. 128 references, 33 figures, 13 tables.

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory

International Nuclear Information System (INIS)

Robinson, Bruce Alan; Funk, David John; Stevens, Patrice Ann

2016-01-01

This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop®1, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options, ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a prescreening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented.

Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-17

This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop®1, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options, ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a prescreening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented.

Comparison of slagging pyrolysis and molten salt incinerators for treating TRU waste at the INEL

International Nuclear Information System (INIS)

1977-11-01

For the comparison, it is assumed that the waste product is required to meet the acceptance criteria of the Waste Isolation Pilot Plant, i.e., low leachability. Slagging pyrolysis incinerates combustible waste and melts noncombustible waste; the resulting slag forms a glass of low leachability. In the molten salt incinerator, combustion occurs at low temperatures with no accumulation of explosive gases, but the waste must have been previously sorted into combustibles and noncombustibles and then shredded. The economics, safety, and technical features are compared. Advantages, disadvantages, and areas of technical uncertainty of the two systems are listed. Development costs and schedules for the two types of incinerators are discussed

Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

International Nuclear Information System (INIS)

Fallis, S.M.

1973-12-01

With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr 90 , Cs 137 , and Pu 239 . Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150 0 C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated

Stabilization/Solidification of Radioactive LiCl-KCl Waste Salt by Using SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}(SAP) Inorganic Composite: Part 2. The Effect of SAP Composition on Stabilization/Solidification

Energy Technology Data Exchange (ETDEWEB)

Ahn, Soo Na; Park, Hwan Seo; Cho, In Hak; Kim, In Tae; Cho, Yong Zun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-03-15

Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of Fe{sub 2}O{sub 3} into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP(Fe=0.1). The experimental results indicated that the addition of Fe{sub 2}O{sub 3} increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing B{sub 2}O{sub 3} into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3 - 4 g of wasteform for final disposal. The final volume would be about 3 - 4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

Medical and Biohazardous Waste Generator's Guide (Revision 2)

International Nuclear Information System (INIS)

Waste Management Group

2006-01-01

These guidelines describe procedures to comply with all Federal and State laws and regulations and Lawrence Berkeley National Laboratory (LBNL) policy applicable to State-regulated medical and unregulated, but biohazardous, waste (medical/biohazardous waste). These guidelines apply to all LBNL personnel who: (1) generate and/or store medical/biohazardous waste, (2) supervise personnel who generate medical/biohazardous waste, or (3) manage a medical/biohazardous waste pickup location. Personnel generating biohazardous waste at the Joint Genome Institute/Production Genomics Facility (JGI/PGF) are referred to the guidelines contained in Section 9. Section 9 is the only part of these guidelines that apply to JGI/PGF. Medical/biohazardous waste referred to in this Web site includes biohazardous, sharps, pathological and liquid waste. Procedures for proper storage and disposal are summarized in the Solid Medical/Biohazardous Waste Disposal Procedures Chart. Contact the Waste Management Group at 486-7663 if you have any questions regarding medical/biohazardous waste management

Charging for waste motivates generators to optimize waste control at the source

International Nuclear Information System (INIS)

Berry, J.B.; Homan, F.J.

1988-01-01

The Department of Energy (DOE) has recognized the need for waste management that incorporates improved waste-handling techniques and more stringent regulatory requirements to prevent future liabilities such as Superfund sites. DOE-Oak Ridge Operations (DOE-ORO) has recognized that an effective waste management program focuses on control at the source and that the burden for responsible waste management can be placed on generators by charging for waste management costs. The principle of including the waste management costs in the total cost of the product, even when the product is research and development, is being implemented at Oak Ridge National Laboratory (ORNL). Charging waste management costs to generators creates an incentive to optimize processes so that less waste is produced, and it provides a basis for determining the cost effectiveness of capital improvements so that the mature phase of waste management can be attained. Improving waste management practices requires a long-range commitment and consistent administration. Making this commitment and providing adequate funding for proper waste disposal are most cost-effective measures than the alternative of paying for remedial actions after improper disposal. This paper summarizes a plan to charge waste generators, the administrative structure of the plan, a comparison between the rate structure and changes in waste disposal operations, and issues that have surfaced as the plan is implemented

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.

Power generation from waste wood

Energy Technology Data Exchange (ETDEWEB)

Nitsche, H

1980-04-18

Since the energy crisis, power generation from waste wood has become increasingly important. The most profitable way to use waste wood in woodworking plants with an annual production of 100 to 150,000 m/sup 3/ solid measure of wood chips and bark is by combustion and thermal energy recovery. In plants with an annual production of 10,000 m/sup 3/ solid measure of wood chips and bark, electric power generation is a suitable application.

Mass transport in bedded salt and salt interbeds

International Nuclear Information System (INIS)

Hwang, Y.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

1989-08-01

Salt is the proposed host rock for geologic repositories of nuclear waste in several nations because it is nearly dry and probably impermeable. Although experiments and experience at potential salt sites indicate that salt may contain brine, the low porosity, creep, and permeability of salt make it still a good choice for geologic isolation. In this paper we summarize several mass-transfer and transport analyses of salt repositories. The mathematical details are given in our technical reports

Conceptual design of retrieval systems for emplaced transuranic waste containers in a salt bed depository. Final report

International Nuclear Information System (INIS)

Fogleman, S.F.

1980-04-01

The US Department of Energy and the Nuclear Regulatory Commission have jurisdiction over the nuclear waste management program. Design studies were previously made of proposed repository site configurations for the receiving, processing, and storage of nuclear wastes. However, these studies did not provide operational designs that were suitable for highly reliable TRU retrieval in the deep geologic salt environment for the required 60-year period. The purpose of this report is to develop a conceptual design of a baseline retrieval system for emplaced transuranic waste containers in a salt bed depository. The conceptual design is to serve as a working model for the analysis of the performance available from the current state-of-the-art equipment and systems. Suggested regulations would be based upon the results of the performance analyses

Geochemical processes in marine salt deposits: Their significance and their implications in connection with disposal of radioactive waste within salt domes

Energy Technology Data Exchange (ETDEWEB)

Herrmann, A G [Goettingen Univ. (Germany, F.R.). Geochemisches Inst.

1980-01-01

Attempts to effect permanent disposal of radioactive wastes in marine evaporites should do nothing to disturb, either in the short or the long term, the present relative stability of such bodies of rock. It is necessary to take account of all of the geochemical and physico-chemical reactions known to have been involved in the processes which formed the evaporites before proceeding to an acceptable strategy for disposal of radionucleides. These processes can be represented as three kinds of metamorphism: 1. solution metamorphism, 2. thermal metamorphism, 3. dynamic metamorphism. In all of the evaporite occurrences in Germany such processes have been influential in altering, on occasion significantly, the primary mineralogical composition and have also promoted a considerable degree of transposition of material. Given similar geochemical and physico-chemical premises, these metamorphic processes could become effective now or in the future. It is therefore necessary to discuss the following criteria when examining salt domes as permanent repositories of highly radioactive substances: (1) Temperatures <= 90/sup 0/ +- 10/sup 0/C at the contact between waste containers and rock salt; (2) Temperatures <= 75/sup 0/C within zones of carnallite rocks; (3) Immobilisation of high-level waste in crystalline forms whenever possible; (4) Systems of additional safety barriers around the waste containers or the unreprocessed spent fuel elements. The geochemical and physical effectiveness of the barriers within an evaporite environment must be guaranteed. For example: Ni-Ti-alloys, corundum, ceramic, anhydrite.

Waste package reference conceptual designs for a repository in salt

International Nuclear Information System (INIS)

1986-02-01

This report provides the reference conceptual waste package designs for the Office of Nuclear Waste Isolation to baseline these designs, thereby establishing the configuration and interface controls necessary, within the Civilian Radioactive Waste Management Program, formerly the National Waste Terminal Storage Program, to proceed in an orderly manner with preliminary design. Included are designs for the current reference defense high-level waste form from the Savannah River Plant, an optimized commercial high-level waste form, and spent fuel which has been disassembled and compacted into a circular bundle containing either 12 pressurized-water reactor or 30 boiling-water reactor assemblies. For compacted spent fuel, it appears economically attractive to standardize the waste package diameter for all fuel types. The reference waste packages consist of the containerized waste form, a low carbon steel overpack, and, after emplacement, a cover of salt. The overpack is a hollow cylinder with a flat head welded to each end. Its design thickness is the sum of the structural thickness required to resist the 15.4-MPa lithostatic pressure plus the corrosion allowance necessary to assure the required structural thickness will exist through the 1000-year containment period. Based on available data and completed analyses, the reference concepts described in this report satisfy all requirements of the US Department of Energy and the US Nuclear Regulatory Commission with reasonable assurance. In addition, sufficient design maturity exists to form a basis for preliminary design; these concepts can be brought under configuration control to serve as reference package designs. Development programs are identified that will be required to support these designs during the licensing process. 19 refs., 37 figs., 31 tabs

INDIVIDUAL DOSIMETRY IN DISPOSAL REPOSITORY OF HEAT-GENERATING NUCLEAR WASTE.

Science.gov (United States)

Pang, Bo; Saurí Suárez, Héctor; Becker, Frank

2016-09-01

Certain working scenarios in a disposal facility of heat-generating nuclear waste might lead to an enhanced level of radiation exposure for workers in such facilities. Hence, a realistic estimation of the personal dose during individual working scenarios is desired. In this study, the general-purpose Monte Carlo N-Particle code MCNP6 (Pelowitz, D. B. (ed). MCNP6 user manual LA-CP-13-00634, Rev. 0 (2013)) was applied to simulate a representative radiation field in a disposal facility. A tool to estimate the personal dose was then proposed by taking into account the influence of individual motion sequences during working scenarios. As basis for this approach, a movable whole-body phantom was developed to describe individual body gestures of the workers during motion sequences. In this study, the proposed method was applied to the German concept of geological disposal in rock salt. The feasibility of the proposed approach was demonstrated with an example of working scenario in an emplacement drift of a rock salt mine. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Identifying suitable piercement salt domes for nuclear waste storage sites

International Nuclear Information System (INIS)

Kehle, R.; e.

1980-08-01

Piercement salt domes of the northern interior salt basins of the Gulf of Mexico are being considered as permanent storage sites for both nuclear and chemically toxic wastes. The suitable domes are stable and inactive, having reached their final evolutionary configuration at least 30 million years ago. They are buried to depths far below the level to which erosion will penetrate during the prescribed storage period and are not subject to possible future reactivation. The salt cores of these domes are themselves impermeable, permitting neither the entry nor exit of ground water or other unwanted materials. In part, a stable dome may be recognized by its present geometric configuration, but conclusive proof depends on establishing its evolutionary state. The evolutionary state of a dome is obtained by reconstructing the growth history of the dome as revealed by the configuration of sedimentary strata in a large area (commonly 3,000 square miles or more) surrounding the dome. A high quality, multifold CDP reflection seismic profile across a candidate dome will provide much of the necessary information when integrated with available subsurface control. Additional seismic profiles may be required to confirm an apparent configuration of the surrounding strata and an interpreted evolutionary history. High frequency seismic data collected in the near vicinity of a dome are also needed as a supplement to the CDP data to permit accurate depiction of the configuration of shallow strata. Such data must be tied to shallow drill hole control to confirm the geologic age at which dome growth ceased. If it is determined that a dome reached a terminal configuration many millions of years ago, such a dome is incapable of reactivation and thus constitutes a stable storage site for nuclear wastes

Technical review of Molten Salt Oxidation

International Nuclear Information System (INIS)

1993-12-01

The process was reviewed for destruction of mixed low-level radioactive waste. Results: extensive development work and scaleup has been documented on coal gasification and hazardous waste which forms a strong experience base for this MSO process; it is clearly applicable to DOE wastes such as organic liquids and low-ash wastes. It also has potential for processing difficult-to-treat wastes such as nuclear grade graphite and TBP, and it may be suitable for other problem waste streams such as sodium metal. MSO operating systems may be constructed in relatively small units for small quantity generators. Public perceptions could be favorable if acceptable performance data are presented fairly; MSO will likely require compliance with regulations for incineration. Use of MSO for offgas treatment may be complicated by salt carryover. Figs, tabs, refs

Newly Generated Liquid Waste Processing Alternatives Study, Volume 1

Energy Technology Data Exchange (ETDEWEB)

Landman, William Henry; Bates, Steven Odum; Bonnema, Bruce Edward; Palmer, Stanley Leland; Podgorney, Anna Kristine; Walsh, Stephanie

2002-09-01

This report identifies and evaluates three options for treating newly generated liquid waste at the Idaho Nuclear Technology and Engineering Center of the Idaho National Engineering and Environmental Laboratory. The three options are: (a) treat the waste using processing facilities designed for treating sodium-bearing waste, (b) treat the waste using subcontractor-supplied mobile systems, or (c) treat the waste using a special facility designed and constructed for that purpose. In studying these options, engineers concluded that the best approach is to store the newly generated liquid waste until a sodium-bearing waste treatment facility is available and then to co-process the stored inventory of the newly generated waste with the sodium-bearing waste. After the sodium-bearing waste facility completes its mission, two paths are available. The newly generated liquid waste could be treated using the subcontractor-supplied system or the sodium-bearing waste facility or a portion of it. The final decision depends on the design of the sodium-bearing waste treatment facility, which will be completed in coming years.

Radioactive Waste Disposal Pilot Plant concept for a New Mexico site

International Nuclear Information System (INIS)

Weart, W.D.

1976-01-01

Twenty years of investigation have shown that disposal of nuclear wastes in deep salt formations is the surest means of isolating these wastes from the biosphere for the extremely long period of time required. A large scale demonstration of this capability will soon be provided by a Radioactive Waste Disposal Pilot Plant (RWDPP) to be developed in southeastern New Mexico. Initially, the pilot plant will accept only ERDA generated waste; high level waste from the commercial power reactor fuel cycle will eventually be accommodated in the pilot plant and the initial RWDPP design will be compatible with this waste form. Selection of a specific site and salt horizon will be completed in June 1976. Conceptual design of the RWDPP and assessment of its environmental impact will be completed by June 1977. Construction is expected to start in 1978 with first waste accepted in 1982. The present concept develops disposal areas for all nuclear waste types in a single salt horizon about 800 meters deep. This single level can accommodate all low level and high level waste generated in the United States through the year 2010. A major constraint on the RWDPP design is the ERDA requirement that all waste be ''readily'' retrievable during the duration of pilot plant operation

Nuclear waste repository simulation experiments, Asse salt mine, Federal Republic of Germany. Annual report 1984

International Nuclear Information System (INIS)

Rothfuchs, T.; Feddersen, H.K.; Schwarzianeck, P.; Staupendahl, G.; Coyle, A.J.; Eckert, J.; Kalia, H.

1986-07-01

This is the second joint annual report (1984) on experiments simulating a nuclear waste repository at the 800-m (2624-ft) level of the Asse salt mine in the Federal Republic of Germany. This report describes the Asse salt mine, the test equipment, and the pretest properties of the salt in the mine and in the vicinity of the test area. Also included are test data for the first 19 months of operation on the following: brine migration rates, thermal mechanical behavior of the salt (including room closure, stress reading, and thermal profiles), and borehole gas pressures. In addition to field data, laboratory analyses of results are included in this report. The duration of the experiment will be 2 years, ending in December 1985

From cerebral salt wasting to diabetes insipidus with adipsia: case report of a child with craniopharyngioma.

Science.gov (United States)

Raghunathan, Veena; Dhaliwal, Maninder Singh; Gupta, Aditya; Jevalikar, Ganesh

2015-03-01

Craniopharyngioma is associated with a wide and interesting variety of sodium states both by itself and following surgical resection. These are often challenging to diagnose, especially given their dynamic nature during the perioperative course. We present the case of a boy with craniopharyngioma who had hyponatremia due to cerebral salt wasting preoperatively, developed diabetes insipidus (DI) intraoperatively and proceeded to develop hypernatremia with adipsic DI. Cerebral salt wasting is a rare presenting feature of craniopharyngioma. Postoperative DI can be associated with thirst abnormalities including adipsia due to hypothalamic damage; careful monitoring and a high index of suspicion are required for its detection. Adipsic DI is a difficult condition to manage; hence a conservative surgical approach is suggested.

Surface displacements and pillar stresses associated with nuclear waste disposal in salt

International Nuclear Information System (INIS)

Hardy, M.P.; St John, C.M.

1977-01-01

A numerical model for regional analysis of stresses and displacement, resulting from heat generating waste placement in underground salt excavations, is presented. The model, which is an extension of that described by McClain and Starfield (1971), is based upon the displacement discontinuity method of stress analysis. It incorporates an empirical characterization of creep behavior of material on the excavation horizon and accounts for thermally induced stresses and displacements. The versatility of this approach is illustrated by the results of three relatively short simulations of test scale disposal facilities at shallow and greater depths. In addition, a three-dimensional code was used to evaluate the surface displacement history for a full-scale repository. This latter code, a thermoelastic analysis, gives an upper bound for the surface movements. It is concluded that the pillar stresses are the result of a complex non-linear interaction of many variables, and the maximum pillar stress can reach several multiples of the tributory-area pillar stress

Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-May 1986

International Nuclear Information System (INIS)

1986-10-01

DOE/CH/10140-05 is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-December 1986

International Nuclear Information System (INIS)

1987-06-01

This document is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

Comparison of the salt domes Asse and Gorleben with regard to their suitability for the final storage of radoactive wastes

International Nuclear Information System (INIS)

Deisenroth, Norbert; Kokorsch, Rudolf

2012-01-01

In Germany, the search for a proper solution to the issue of final disposal of radioactive wastes is complicated by political leaders. The Gorleben moratorium from October 2000 delayed the proper solution unnecessary to ten years. Asse proves that salt domes such as Gorleben do not offer a permanent partitioning of the waste over the biosphere. With this in mind, the authors of the contribution under consideration compare the two salt domes Gorleben and Asse from a mining and geological point of view based on publicly available data with regard to their suitability for the disposal of radioactive waste.

Worth its salt?

Science.gov (United States)

The idea that all underground salt deposits can serve as storage sites for toxic and nuclear waste does not always hold water—literally. According to Daniel Ronen and Brian Berkowitz of Israel's Weizmann Institute of Science and Yoseph Yechieli of the Geological Survey of Israel, some buried salt layers are in fact highly conductive of liquids, suggesting that wastes buried in their confines could easily leech into groundwater and nearby soil.When drilling three wells into a 10,000-year-old salt layer near the Dead Sea, the researchers found that groundwater had seeped into the layer and had absorbed some of its salt.

Preliminary area selection considerations for radioactive waste repositories in bedded salt

International Nuclear Information System (INIS)

Wagoner, J.L.; Steinborn, T.L.

1979-01-01

This guide describes an approach to selection of areas of bedded salt which contain potentially suitable sites for the storage of radioactive waste. To evaluate a site selected by a license applicant, it is necessary to understand the technical site characteristics which should be considered in the preliminary phase of site selection. These site characteristics are presented here in checklist form, and each item is accompanied by a discussion which explains its significance. These qualitative considerations are used first to select an area of interest within a broad geologic or geomorphic region. Once an area has been selected, more quantitative information must be acquired to determine whether the proposed site meets the resultations for storage of nuclear waste

Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

Energy Technology Data Exchange (ETDEWEB)

Fallis, S.M.

1973-12-01

With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr/sup 90/, Cs/sup 137/, and Pu/sup 239/. Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150/sup 0/C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated.

Estimation of restaurant solid waste generation rates

International Nuclear Information System (INIS)

Heck, H.H.; Major, I.

2002-01-01

Most solid waste utilities try to create a billing schedule that is proportional to solid waste generation rates. This research was trying to determine if the current billing rate structure was appropriate or if a different rate structure should be implemented. A multiple regression model with forward stepwise addition was developed which accurately predicts weekly solid waste generation rates for restaurants. The model was based on a study of daily solid waste generation at twenty-one different businesses. The weight and volume of solid waste generated was measure daily for two weeks during the winter and two weeks during the summer. Researchers followed the collection truck and measured the volume and weight of the container contents. Data was collected on the following independent variables describing each establishment; weight of waste per collection, volume per collection, container utilization factor, building area, contract haulers bill, yearly property tax, yearly solid waste tax, average number of collections per week, type of restaurant, modal number of collections per week, storage container size, waste density, number of employees, number of hours open per week, and weekly collection capacity (collections per week times storage container size). Independent variables were added to the regression equation based on their partial correlation coefficient and confidence level. The regression equations developed had correlation coefficients of 0.87 to 1.00, which was much better than the correlation coefficient (0.84) of an existing model DeGeare and Ongerth (1971) and a correlation coefficient of 0.54 based on the current solid waste disposal tax. (author)

Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment

International Nuclear Information System (INIS)

Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

1983-10-01

Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables

Corrosion of carbon steel in saturated high-level waste salt solutions

International Nuclear Information System (INIS)

Wiersma, B.J.; Parish, W.R.

1997-01-01

High level waste stored as crystallized salts is to be removed from carbon steel tanks by water dissolution. Dissolution of the saltcake must be performed in a manner which will not impact the integrity of the tank. Corrosion testing was performed to determine the amount of corrosion inhibitor that must be added to the dissolution water in order to ensure that the salt solution formed would not induce corrosion degradation of the tank materials. The corrosion testing performed included controlled potential slow strain rate, coupon immersion, and potentiodynamic polarization tests. These tests were utilized to investigate the susceptibility of the cooling coil material to stress corrosion cracking in the anticipated environments. No evidence of SCC was observed in any of the tests. Based on these results, the recommended corrosion requirements were that the temperature of the salt solution be less than 50 degrees C and that the minimum hydroxide concentration be 0.4 molar. It was also recommended that the hydroxide concentration not stay below 0.4 molar for longer than 45 days

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

International Nuclear Information System (INIS)

Ramthun, H.

1981-08-01

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

Factors determining waste generation in Spanish towns and cities.

Science.gov (United States)

Prades, Miriam; Gallardo, Antonio; Ibàñez, Maria Victoria

2015-01-01

This paper analyzes the generation and composition of municipal solid waste in Spanish towns and cities with more than 5000 inhabitants, which altogether account for 87% of the Spanish population. To do so, the total composition and generation of municipal solid waste fractions were obtained from 135 towns and cities. Homogeneity tests revealed heterogeneity in the proportions of municipal solid waste fractions from one city to another. Statistical analyses identified significant differences in the generation of glass in cities of different sizes and in the generation of all fractions depending on the hydrographic area. Finally, linear regression models and residuals analysis were applied to analyze the effect of different demographic, geographic, and socioeconomic variables on the generation of waste fractions. The conclusions show that more densely populated towns, a hydrographic area, and cities with over 50,000 inhabitants have higher waste generation rates, while certain socioeconomic variables (people/car) decrease that generation. Other socioeconomic variables (foreigners and unemployment) show a positive and null influence on that waste generation, respectively.

Second Generation Waste Package Design Study

International Nuclear Information System (INIS)

Armijo, J.S.; Misra, M.; Kar, Piyush

2007-01-01

The following describes the objectives of Project Activity 023 ''Second Generation Waste Package Design Study'' under DOE Cooperative Agreement DE-FC28-04RW12232. The objectives of this activity are: to review the current YMP baseline environment and establish corrosion test environments representative of the range of dry to intermittently wet conditions expected in the drifts as a function of time; to demonstrate the oxidation and corrosion resistance of A588 weathering steel and reference Alloy 22 samples in the representative dry to intermittently dry conditions; and to evaluate backfill and design features to improve the thermal performance analyses of the proposed second-generation waste packages using existing models developed at the University of Nevada, Reno(UNR). The work plan for this project activity consists of three major tasks: Task 1. Definition of expected worst-case environments (humidity, liquid composition and temperature) at waste package outer surfaces as a function of time, and comparison with environments defined in the YMP baseline; Task 2. Oxidation and corrosion tests of proposed second-generation outer container material; and Task 3. Second Generation waste package thermal analyses. Full funding was not provided for this project activity

Study on reducing the generation of general waste

International Nuclear Information System (INIS)

Takahashi, Toshio; Aoki, Isao; Watahiki, Masatoshi

2000-04-01

On August 1999, the Director of Tokai Work proposed an activity regarding recycling and reuse of general waste generated from Tokai Works. The activity was initiated by the Waste Management and Fuel Cycle Research Center, and is now being in progress through out the Tokai Works. In the course of this activity, Plutonium Fuel Center had settled the working Group and the issues related to the waste reductive have been examined. This report collects the problems that became obvious through the survey of existing segregation method, treatment process, and the amount of the waste generation, and accounts for the concrete methodology for the recycling and reuse of general waste. In order to reduce waste, it is necessary to aware of the facing issues and adopt the countermeasures proposed in this report whenever possible. The activity will then leads us to reduce waste generation, which in turn will enable us to make 100% waste recycling possible. (author)

Tetraphenylborate Solubility in High Ionic Strength Salt Solutions

International Nuclear Information System (INIS)

Serkiz, S.M.; Ginn, J.D.; Jurgensen, A.R.

1998-04-01

Solubility of sodium and potassium salts of the tetraphenylborate ion (TPB) in simulated Savannah River Site High Level Waste was investigated. Data generated from this study allow more accurate predictions of TPB solubility at the In-Tank Precipitation (ITP) facility. Because previous research showed large deviations in the observed solubility of TPB salts when compared with model predictions, additional data were generated to better understand the solubility of TPB in more complex systems of high ionic strength and those containing both potassium and sodium. These data allow evaluation of the ability of current models to accurately predict equilibrium TPB concentrations over the range of experimental conditions investigated in this study

Phase Equilibrium Studies of Savannah River Tanks and Feed Streams for the Salt Waste Processing Facility

Energy Technology Data Exchange (ETDEWEB)

Weber, C.F.

2001-06-19

A chemical equilibrium model is developed and used to evaluate supersaturation of tanks and proposed feed streams to the Salt Waste Processing Facility. The model uses Pitzer's model for activity coefficients and is validated by comparison with a variety of thermodynamic data. The model assesses the supersaturation of 13 tanks at the Savannah River Site (SRS), indicating that small amounts of gibbsite and or aluminosilicate may form. The model is also used to evaluate proposed feed streams to the Salt Waste Processing Facility for 13 years of operation. Results indicate that dilutions using 3-4 M NaOH (about 0.3-0.4 L caustic per kg feed solution) should avoid precipitation and reduce the Na{sup +} ion concentration to 5.6 M.

In-situ stress measurements - results of experiments performed at the ASSE salt mine - Federal Republic of Germany

International Nuclear Information System (INIS)

Feddersen, H.K.

1989-01-01

High-level nuclear wastes are heat generating wastes. Heat will be transferred to the surrounding salt formation. This heating of the host rock will result in an increased temperature and in stress changes. From 1983 through 1985 two underground tests were conducted in the Asse Salt Mine (Federal Republic of Germany) in which, among others, thermally induced stress changes were investigated. These tests are discussed in this paper

Just-in-time characterization and certification of DOE-generated wastes

Energy Technology Data Exchange (ETDEWEB)

Arrenholz, D.A.; Primozic, F.J. [Benchmark Environmental Corp., Albuquerque, NM (United States); Robinson, M.A. [Los Alamos National Lab., NM (United States)

1995-12-31

Transportation and disposal of wastes generated by Department of Energy (DOE) activities, including weapons production and decontamination and decommissioning (D&D) of facilities, require that wastes be certified as complying with various regulations and requirements. These certification requirements are typically summarized by disposal sites in their specific waste acceptance criteria. Although a large volume of DOE wastes have been generated by past activities and are presently in storage awaiting disposal, a significant volume of DOE wastes, particularly from D&D projects. have not yet been generated. To prepare DOE-generated wastes for disposal in an efficient manner. it is suggested that a program of just-in-time characterization and certification be adopted. The goal of just-in-time characterization and certification, which is based on the just-in-time manufacturing process, is to streamline the certification process by eliminating redundant layers of oversight and establishing pro-active waste management controls. Just-in-time characterization and certification would rely on a waste management system in which wastes are characterized at the point of generation, precertified as they are generated (i.e., without iterative inspections and tests subsequent to generation and storage), and certified at the point of shipment, ideally the loading dock of the building from which the wastes are generated. Waste storage would be limited to accumulating containers for cost-efficient transportation.

Just-in-time characterization and certification of DOE-generated wastes

International Nuclear Information System (INIS)

Arrenholz, D.A.; Primozic, F.J.; Robinson, M.A.

1995-01-01

Transportation and disposal of wastes generated by Department of Energy (DOE) activities, including weapons production and decontamination and decommissioning (D ampersand D) of facilities, require that wastes be certified as complying with various regulations and requirements. These certification requirements are typically summarized by disposal sites in their specific waste acceptance criteria. Although a large volume of DOE wastes have been generated by past activities and are presently in storage awaiting disposal, a significant volume of DOE wastes, particularly from D ampersand D projects. have not yet been generated. To prepare DOE-generated wastes for disposal in an efficient manner. it is suggested that a program of just-in-time characterization and certification be adopted. The goal of just-in-time characterization and certification, which is based on the just-in-time manufacturing process, is to streamline the certification process by eliminating redundant layers of oversight and establishing pro-active waste management controls. Just-in-time characterization and certification would rely on a waste management system in which wastes are characterized at the point of generation, precertified as they are generated (i.e., without iterative inspections and tests subsequent to generation and storage), and certified at the point of shipment, ideally the loading dock of the building from which the wastes are generated. Waste storage would be limited to accumulating containers for cost-efficient transportation

Radioactive waste assessment using 'minimum waste generation' scenario - summary report March 1984

International Nuclear Information System (INIS)

Richardson, J.A.; Goodill, D.R.; Tymons, B.J.

1984-11-01

This report describes an assessment of radioactive waste management arisings from a defined nuclear power generation - Scheme 1. Scheme 1 assumes a minimum waste generation scenario with raw waste arisings from 3 main groups; (i) existing and committed commercial reactors; (ii) fuel reprocessing plants, (iii) research, industry and medicine. No decommissioning wastes are considered except for arisings from the final fuel cores from decommissioned reactors. The study uses the SIMULATION2 code which models waste material flows through the system. With a knowledge of the accumulations and average production rates of the raw wastes and their isotopic compositions (or total activities), the rates at which conditioned wastes become available for transportation and disposal are calculated, with specific activity levels. The data bases for the inventory calculations and the assumptions concerning future operation of nuclear facilities were those current in 1983. Both the inventory data and plans for the future of existing nuclear installations have been updated since these calculations were completed. Therefore the results from this assessment do not represent the most up-to-date information available. The report does, however, illustrate the methodology of assessment and indicates the type of information that can be generated. (author)

Waste minimization for commercial radioactive materials users generating low-level radioactive waste

International Nuclear Information System (INIS)

Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S.; Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L.

1991-07-01

The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations. This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature

Solid waste generation in reprocessing nuclear fuel

International Nuclear Information System (INIS)

North, E.D.

1975-01-01

Estimates are made of the solid wastes generated annually from a 750-ton/year plant (such as the NFS West Valley plant): high-level waste, hulls, intermediate level waste, failed equipment, HEPA filters, spent solvent, alpha contaminated combustible waste, and low specific activity waste. The annual volume of each category is plotted versus the activity level

Radiotoxicity of nuclear waste and disposal possibilities in the Netherlands

International Nuclear Information System (INIS)

Weber, J.

1980-01-01

A description is given of the evaluation of the long-term risks of the disposal of nuclear waste in salt-domes. It is generic for the salt-domes in the N-E of the Netherlands. In the far future the isolation of the waste from the biosphere provided by the salt can be violated by diapyrism and by dissolution. Although improbable a diapyrism of 2.5 mm/year cannot be excluded. As this pathway gives the higher risks for future generations it was taken as the basis for the evaluation. (Auth.)

Evaluation of bituminized waste reactivity

International Nuclear Information System (INIS)

Camaro, S.; Moulinier, D.

2000-01-01

The bituminization process has been used for conditioning low and medium level (LML) radioactive waste, particularly to immobilize coprecipitation slurries and evaporation concentrates generated by effluent treatment. The process consists in mixing bitumen matrix with inactive soluble and slightly soluble salts added to insolubilize the radionuclides or resulting from the neutralization of acid effluents. This operation is performed at a sufficient temperature - depending on waste composition and bitumen grade to ensure the flow of the resulting mixture into metal containers. Exothermicity due to salts/salts or salts/bitumen reactions depending on the type of waste can be induced during or after the mixing step. This could produce an additional heat emission that the drum must be able to release to avoid a potentially incidental pattern with ignition risk, explaining why the CEA has been involved in evaluating the thermal reactivity of bituminized waste and its repercussions on the bituminization process. Given the difficulty of discriminating each exothermal reaction, the characterization of a global reactivity appears as a further precautionary measure, in addition to the definition of a working safety margin. The CEA has accordingly developed studies on this aspect. The article discusses the experimental methodology developed for the determination of the global reactivity. (authors)

A Survey of Municipal Solid Waste Generation in 22 Regions of Tehran With Solid Waste Reduction Approach

Directory of Open Access Journals (Sweden)

MA Abduli

2015-07-01

Methods: The study was a descriptive cross-sectional one conducted from 2010 to 2014. Relevant officials of the waste recovery in 22 regions of Tehran were approached in order to collect data about municipal solid waste generation through interviewing, filling out questionnaires, conducting field visits from Aradkooh Disposal and Processing Complex and collecting information on disposal and destiny of wastes. Then the data were compiled and analyzed. Results: Total solid waste generation in Tehran from 2010 to 2014 amounted to respectively 3389662, 3399344, 3449338 and 3245157 Metric Tons, categorized into three groups of municipal, companies and townships and hospital wastes. Most of the generated waste produced in Tehran was that of households and commercial (known as municipal waste from 22 Regions of Tehran. Based on the surveys conducted, per capita solid waste generation of 11 regions of Tehran ranged from 550 to 1000 grams and in other 11 ones from 1000 to 1521 grams per capita per day. The lowest and highest waste generation rate belonged respectively to region 13 with 556 grams and region 12 with 1521 grams per capita per day in 2011. Conclusion: Comparing per capita generation of municipal solid waste in different municipal regions in Tehran with maximum acceptable capacity of waste generation indicates the deviation of waste generation of all Tehran regions from the standard acceptable amount. Therefore, not only is it necessary to plan and take strategic measures to reduce Tehran waste generation but also these programs and measures should be specific to each region considering its specifications and solid waste quality and quantity.

Evaluation of iron-base materials for waste package containers in a salt repository

International Nuclear Information System (INIS)

Westerman, R.E.; Nelson, J.L.; Kuhn, W.L.; Basham, S.G.; Moak, D.A.; Pitman, S.G.

1983-11-01

Design studies for high-level nuclear waste packages for salt repositories have identified low-carbon steel as a candidate material for containers. Among the requirements are strength, corrosion resistance, and fabricability. The studies of the corrosion resistance and structural stability of iron-base materials (particularly low-carbon steel) are treated in this paper. The materials have been exposed in brines that are characteristic of the potential sites for salt repositories. The effects of temperature, radiation level, oxygen level and other parameters are under investigation. The initial development of corrosion models for these environments is presented with discussion of the key mechanisms under consideration. 6 references, 5 figures

Electrodialysis Treatment of Industrial Waste Waters That Generaters Little Sludge; Tratamiento de aguas residuales industriales por electrodialisis, tecnica que genera pocos lodos

Energy Technology Data Exchange (ETDEWEB)

Savastoma Mariel, C. [Universidad de Cantabria. Santander (Spain)

2002-07-01

Electrodialysis with bipolar membranes is reviewed as a separation technique and details are shown. The method is used in several process applications but its adoption looks particularly attractive in the treatment of waste water streams consisting primarily of diluted residual salts and electrolytes. It is argued that the main advantage shown by the technology compared to the classical treatment consisting of adding lime to neutralize the waste waters resides in tis avoiding the generation of voluminous sludges thus permitting savings in the management and disposal of the mud. Other interesting applications are those operations where generating acidity and/or alkalinity with addition of no extra chemicals is desirable, such as the regeneration of ion exchange resins. (Author) 34 refs.

Solid waste generation and characterization in the University of Lagos for a sustainable waste management.

Science.gov (United States)

Adeniran, A E; Nubi, A T; Adelopo, A O

2017-09-01

Waste characterization is the first step to any successful waste management policy. In this paper, the characterization and the trend of solid waste generated in University of Lagos, Nigeria was carried out using ASTM D5231-92 and Resource Conservation Reservation Authority RCRA Waste Sampling Draft Technical Guidance methods. The recyclable potential of the waste is very high constituting about 75% of the total waste generated. The estimated average daily solid waste generation in Unilag Akoka campus was estimated to be 32.2tons. The solid waste characterization was found to be: polythene bags 24% (7.73tons/day), paper 15% (4.83tons/day), organic matters 15%, (4.83tons/day), plastic 9% (2.90tons/day), inert materials 8% (2.58tons/day), sanitary 7% (2.25tons/day), textile 7% (2.25tons/day), others 6% (1.93tons/day), leather 4% (1.29tons/day) metals 3% (0.97tons/day), glass 2% (0.64tons/day) and e-waste 0% (0.0tons/day). The volume and distribution of polythene bags generated on campus had a positive significant statistical correlation with the distribution of commercial and academic structures on campus. Waste management options to optimize reuse, recycling and reduce waste generation were discussed. Copyright © 2017 Elsevier Ltd. All rights reserved.

Impact of socioeconomic status on municipal solid waste generation rate.

Science.gov (United States)

Khan, D; Kumar, A; Samadder, S R

2016-03-01

The solid waste generation rate was expected to vary in different socioeconomic groups due to many environmental and social factors. This paper reports the assessment of solid waste generation based on different socioeconomic parameters like education, occupation, income of the family, number of family members etc. A questionnaire survey was conducted in the study area to identify the different socioeconomic groups that may affect the solid waste generation rate and composition. The average waste generated in the municipality is 0.41 kg/capita/day in which the maximum waste was found to be generated by lower middle socioeconomic group (LMSEG) with average waste generation of 0.46 kg/capita/day. Waste characterization indicated that there was no much difference in the composition of wastes among different socioeconomic groups except ash residue and plastic. Ash residue is found to increase as we move lower down the socioeconomic groups with maximum (31%) in lower socioeconomic group (LSEG). The study area is a coal based city hence application of coal and wood as fuel for cooking in the lower socioeconomic group is the reason for high amount of ash content. Plastic waste is maximum (15%) in higher socioeconomic group (HSEG) and minimum (1%) in LSEG. Food waste is a major component of generated waste in almost every socioeconomic group with maximum (38%) in case of HSEG and minimum (28%) in LSEG. This study provides new insights on the role of various socioeconomic parameters on generation of household wastes. Copyright © 2016 Elsevier Ltd. All rights reserved.

Food waste from Danish households: Generation and composition

DEFF Research Database (Denmark)

Edjabou, Maklawe Essonanawe; Petersen, Claus; Scheutz, Charlotte

2016-01-01

Sustainable solutions for reducing food waste require a good understanding of food waste generation and composition, including avoidable and unavoidable food waste. We analysed 12 tonnes of residual household waste collected from 1474 households, without source segregation of organic waste. Food...... waste was divided into six fractions according to avoidability, suitability for home-composting and whether or not it was cooked, prepared or had been served within the household. The results showed that the residual household waste generation rate was 434 ± 18 kg per household per year, of which 183...... ± 10 kg per year was food waste. Unavoidable food waste amounted to 80 ± 6 kg per household per year, and avoidable food waste was 103 ± 9 kg per household per year. Food waste mass was influenced significantly by the number of occupants per household (household size) and the housing type. The results...

Pathways for Disposal of Commercially-Generated Tritiated Waste

Energy Technology Data Exchange (ETDEWEB)

Halverson, Nancy V. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Environmental Sciences and Biotechnology

2016-09-26

From a waste disposal standpoint, tritium is a major challenge. Because it behaves like hydrogen, tritium exchanges readily with hydrogen in the ground water and moves easily through the ground. Land disposal sites must control the tritium activity and mobility of incoming wastes to protect human health and the environment. Consequently, disposal of tritiated low-level wastes is highly regulated and disposal options are limited. The United States has had eight operating commercial facilities licensed for low-level radioactive waste disposal, only four of which are currently receiving waste. Each of these is licensed and regulated by its state. Only two of these sites accept waste from states outside of their specified regional compact. For waste streams that cannot be disposed directly at one of the four active commercial low-level waste disposal facilities, processing facilities offer various forms of tritiated low-level waste processing and treatment, and then transport and dispose of the residuals at a disposal facility. These processing facilities may remove and recycle tritium, reduce waste volume, solidify liquid waste, remove hazardous constituents, or perform a number of additional treatments. Waste brokers also offer many low-level and mixed waste management and transportation services. These services can be especially helpful for small-quantity tritiated-waste generators, such as universities, research institutions, medical facilities, and some industries. The information contained in this report covers general capabilities and requirements for the various disposal/processing facilities and brokerage companies, but is not considered exhaustive. Typically, each facility has extensive waste acceptance criteria and will require a generator to thoroughly characterize their wastes. Then a contractual agreement between the waste generator and the disposal/processing/broker entity must be in place before waste is accepted. Costs for tritiated waste

Pathways for Disposal of Commercially-Generated Tritiated Waste

International Nuclear Information System (INIS)

Halverson, Nancy V.

2016-01-01

From a waste disposal standpoint, tritium is a major challenge. Because it behaves like hydrogen, tritium exchanges readily with hydrogen in the ground water and moves easily through the ground. Land disposal sites must control the tritium activity and mobility of incoming wastes to protect human health and the environment. Consequently, disposal of tritiated low-level wastes is highly regulated and disposal options are limited. The United States has had eight operating commercial facilities licensed for low-level radioactive waste disposal, only four of which are currently receiving waste. Each of these is licensed and regulated by its state. Only two of these sites accept waste from states outside of their specified regional compact. For waste streams that cannot be disposed directly at one of the four active commercial low-level waste disposal facilities, processing facilities offer various forms of tritiated low-level waste processing and treatment, and then transport and dispose of the residuals at a disposal facility. These processing facilities may remove and recycle tritium, reduce waste volume, solidify liquid waste, remove hazardous constituents, or perform a number of additional treatments. Waste brokers also offer many low-level and mixed waste management and transportation services. These services can be especially helpful for small-quantity tritiated-waste generators, such as universities, research institutions, medical facilities, and some industries. The information contained in this report covers general capabilities and requirements for the various disposal/processing facilities and brokerage companies, but is not considered exhaustive. Typically, each facility has extensive waste acceptance criteria and will require a generator to thoroughly characterize their wastes. Then a contractual agreement between the waste generator and the disposal/processing/broker entity must be in place before waste is accepted. Costs for tritiated waste

The dispersal and impact of salt from surface storage piles the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Reith, C.C.; Louderbough, E.T.

1986-01-01

A comprehensive program of ecological studies occurs at the Waste Isolation Pilot Plant (WIPP) in an effort to detect and quantify impacts of excavated salt which is stored on the surface in two piles: one having originated in 1980, the other in 1984. Both piles are surrounded by berms which channel runoff to holding ponds, so nearly all dispersal is due to the resuspension, transport, and deposition of salt particles by wind. Ecological parameters which have been monitored since 1984 include: visual evidence (via photography), soil properties, microbial activity, leaf-litter decomposition, seedling emergence, plant foliar cover, and plant species diversity. These are periodically assessed at experimental plots near the salt piles, and at control plots several kilometers away

SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

S. C. Khamankar

2000-06-20

The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

SITE GENERATED RADIOLOGICAL WASTE HANDLING SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

S. C. Khamankar

2000-01-01

The Site Generated Radiological Waste Handling System handles radioactive waste products that are generated at the geologic repository operations area. The waste is collected, treated if required, packaged for shipment, and shipped to a disposal site. Waste streams include low-level waste (LLW) in solid and liquid forms, as-well-as mixed waste that contains hazardous and radioactive constituents. Liquid LLW is segregated into two streams, non-recyclable and recyclable. The non-recyclable stream may contain detergents or other non-hazardous cleaning agents and is packaged for shipment. The recyclable stream is treated to recycle a large portion of the water while the remaining concentrated waste is packaged for shipment; this greatly reduces the volume of waste requiring disposal. There will be no liquid LLW discharge. Solid LLW consists of wet solids such as ion exchange resins and filter cartridges, as-well-as dry active waste such as tools, protective clothing, and poly bags. Solids will be sorted, volume reduced, and packaged for shipment. The generation of mixed waste at the Monitored Geologic Repository (MGR) is not planned; however, if it does come into existence, it will be collected and packaged for disposal at its point of occurrence, temporarily staged, then shipped to government-approved off-site facilities for disposal. The Site Generated Radiological Waste Handling System has equipment located in both the Waste Treatment Building (WTB) and in the Waste Handling Building (WHB). All types of liquid and solid LLW are processed in the WTB, while wet solid waste from the Pool Water Treatment and Cooling System is packaged where received in the WHB. There is no installed hardware for mixed waste. The Site Generated Radiological Waste Handling System receives waste from locations where water is used for decontamination functions. In most cases the water is piped back to the WTB for processing. The WTB and WHB provide staging areas for storing and shipping LLW

Areal thermal loading recommendations for nuclear waste repositories in salt

International Nuclear Information System (INIS)

Russell, J.E.

1979-06-01

This document gives a wider understanding of the history of the recommended thermal loadings in salt for both high-level waste (HLW) from fresh UO 2 -fueled, light-water reactors (LWR) with no recycle and spent unreprocessed fuel (SURF) from LWRs. Aspects of the current recommendations that need further study are identified. Finally, an interim set of design thermal-loading recommendations are given that have a common rationale of satisfying performance limits within our current state of knowledge. These recommendations are made on a generic rather than a site-specific basis. 11 figures, 5 tables

Waste Minimization Study on Pyrochemical Reprocessing Processes

International Nuclear Information System (INIS)

Boussier, H.; Conocar, O.; Lacquement, J.

2006-01-01

Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal