Evaluation of bitumens for nuclear facilities radioactive waste immobilization

International Nuclear Information System (INIS)

Guzella, Marcia F.R.; Silva, Tania V. da; Loiola, Roberto; Monte, Lauro J.B.

2000-01-01

The activities developed at the Nuclear Technology Development Centre, Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN, include the research and development work of the radioactive wastes immobilization in different kind of bitumen. The present work describes the bituminization of simulated low level wastes of evaporator concentrates.Two types of bitumen are used for incorporation of the simulated wastes generated by nuclear power plants. Studies on rheological properties, leaching data, differential thermoanalysis and water content of the waste-products have been carried out. (author)

Preliminary evaluation of alternative forms for immobilization of Hanford high-level defense wastes

International Nuclear Information System (INIS)

Schulz, W.W.; Beary, M.M.; Gallagher, S.A.; Higley, B.A.; Johnston, R.G.; Jungfleisch, F.M.; Kupfer, M.J.; Palmer, R.A.; Watrous, R.A.; Wolf, G.A.

1980-09-01

A preliminary evaluation of solid waste forms for immobilization of Hanford high-level radioactive defense wastes is presented. Nineteen different waste forms were evaluated and compared to determine their applicability and suitability for immobilization of Hanford salt cake, sludge, and residual liquid. This assessment was structured to address waste forms/processes for several different leave-retrieve long-term Hanford waste management alternatives which give rise to four different generic fractions: (1) sludge plus long-lived radionuclide concentrate from salt cake and residual liquid; (2) blended wastes (salt cake plus sludge plus residual liquid); (3) residual liquid; and (4) radionuclide concentrate from residual liquid. Waste forms were evaluated and ranked on the basis of weighted ratings of seven waste form and seven process characteristics. Borosilicate Glass waste forms, as marbles or monoliths, rank among the first three choices for fixation of all Hanford high-level wastes (HLW). Supergrout Concrete (akin to Oak Ridge National Laboratory Hydrofracture Process concrete) and Bitumen, low-temperature waste forms, rate high for bulk disposal immobilization of high-sodium blended wastes and residual liquid. Certain multi-barrier (e.g., Coated Ceramic) and ceramic (SYNROC Ceramic, Tailored Ceramics, and Supercalcine Ceramic) waste forms, along with Borosilicate Glass, are rated as the most satisfactory forms in which to incorporate sludges and associated radionuclide concentrates. The Sol-Gel process appears superior to other processes for manufacture of a generic ceramic waste form for fixation of Hanford sludge. Appropriate recommendations for further research and development work on top ranking waste forms are made

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

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

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

High level nuclear wastes

International Nuclear Information System (INIS)

Lopez Perez, B.

1987-01-01

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

Remote handling in the Plutonium Immobilization Project: Plutonium conversion and first stage immobilization

International Nuclear Information System (INIS)

Brault, J.R.

2000-01-01

Since the break up of the Soviet Union at the end of the Cold War, the United States and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

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

International Nuclear Information System (INIS)

Burgard, K.C.

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Burgard, K.C.

1998-04-09

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

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

International Nuclear Information System (INIS)

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

1976-06-01

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

Managing the process for storage and disposal of immobilized high- and low-level tank waste at the Hanford Site

International Nuclear Information System (INIS)

Murkowski, R.J.

1998-01-01

Lockheed Martin Hanford Corporation (LMHC) is one of six subcontractors under Fluor Daniel Hanford, Inc., the Management and Integration contractor for the Project Hanford Management Contract working for the US Department of Energy. One of LMHC's responsibilities is to prepare storage and disposal facilities to receive immobilized high and low-level tank waste by June of 2002. The immobilized materials are to be produced by one or more vendors working under a privatization contract. The immobilized low-activity waste is to be permanently disposed of at the Hanford Site while the immobilized high-level waste is to be stored at the Hanford Site while awaiting shipment to the offsite repository. Figure 1 is an overview of the entire cleanup mission with the disposal portion of the mission. Figure 2 is a representation of major activities required to complete the storage and disposal mission. The challenge for the LNIHC team is to understand and plan for accepting materials that are described in the Request for Proposal. Private companies will submit bids based on the Request for Proposal and other Department of Energy requirements. LMHC, however, must maintain sufficient flexibility to accept modifications that may occur during the privatization bid/award process that is expected to be completed by May 1998. Fundamental to this planning is to minimize the risks of stand-by costs if storage and disposal facilities are not available to receive the immobilized waste. LMHC has followed a rigorous process for the identification of the functions and requirements of the storage/disposal facilities. A set of alternatives to meet these functions and requirements were identified and evaluated. The alternatives selected were (1) to modify four vaults for disposal of immobilized low-activity waste, and (2) to retrofit a portion of the Canister Storage Building for storage of immobilized high-level waste

An optimized approach towards the treatment of high level liquid waste in the nuclear cycle

International Nuclear Information System (INIS)

Maio, V.; Todd, T.; Law, J.; Roach, J.; Sabharwall, P.

2006-01-01

Full text: One key long-standing issue that must be overcome to realize the successful growth of nuclear power is an economical, politically acceptable, stakeholder-compatible, and technically feasible resolution pertaining to the safe treatment and disposal of high-level liquid radioactive waste (HLLW). In addition to spent nuclear reactor fuel, HLLW poses a unique challenge in regard to environmental and security concerns, since future scenarios for a next generation of domestic and commercialized nuclear fuel cycle infrastructures must include reprocessing - the primary source of HLLW-to ensure the cost effectiveness of nuclear power as well as mitigate any threats as related to proliferation. Past attempts to immobilize HLLW - generated by both the weapons complex and the commercial power sector-have been plagued by an inability to convince the public and some technical peer reviewers that any proposed geological disposal sites (e.g., Yucca Mountain) can accommodate and contain the HLLW for a period of geological time equivalent to ten fold the radiological half-life of the longest lived of the actinides remaining after reprocessing. The paper explores combined equipment and chemical processing approaches for advancing and economizing the immobilization of high level liquid waste to ensure its long term durability, its decoupling from the unknown behavior of the repository over long geological time periods, and its economical formulation as required for the nuclear fuel cycle of the future. One approach involves the investigation of crystalline based waste forms as opposed to the glass/amorphous based waste forms, and how recent developments in crystalline forms show promise in sequestering the long lived actinides for over tens of millions of years. Another approach -compatible with the first- involves the use of an alternative melter technology-the Cold Crucible Induction Melter (CCIM)- to overcome the engineering material problems of Joule Heated Meters (JHM

Material chemistry challenges in vitrification of high level radioactive waste

International Nuclear Information System (INIS)

Kaushik, C.P.

2008-01-01

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

A review of glass-ceramics for the immobilization of nuclear fuel recycle wastes

International Nuclear Information System (INIS)

Hayward, P.J.

1987-01-01

This report reviews the status of the Canadian, German, U.S., Japanese, U.S.S.R. and Swedish programs for the development of glass-ceramic materials for immobilizing the high-level radioactive wastes arising from the recycling of used nuclear fuel. The progress made in these programs is described, with emphasis on the Canadian program for the development of sphene-based glass-ceramics. The general considerations of product performance and process feasibility for glass-ceramics as a category of waste form material are discussed. 137 refs

Mathematical model for choosing the nuclear safe matrix compositions for fissile material immobilization

International Nuclear Information System (INIS)

Gorshtein, A.I.; Matyunin, Yu.I.; Poluehktov, P.P.

2000-01-01

A mathematical model is proposed for preliminary choice of the nuclear safe matrix compositions for fissile material immobilization. The IBM PC computer software for nuclear safe matrix composition calculations is developed. The limiting concentration of fissile materials in the some used and perspective nuclear safe matrix compositions for radioactive waste immobilization is calculated [ru

Immobilization of defense high-level waste: an assessment of technological strategies and potential regulatory goals. Volume II

International Nuclear Information System (INIS)

1979-06-01

This volume contains the following appendices: selected immobilization processes, directory of selected European organizations involved in HLW management, U.S. high-level waste inventories, and selected European HLW program

Long-term high-level waste technology program

International Nuclear Information System (INIS)

1980-04-01

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

Immobilization of high-level wastes into sintered glass: 1

International Nuclear Information System (INIS)

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

1987-01-01

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

Experimental study on cesium immobilization in struvite structures.

Science.gov (United States)

Wagh, Arun S; Sayenko, S Y; Shkuropatenko, V A; Tarasov, R V; Dykiy, M P; Svitlychniy, Y O; Virych, V D; Ulybkina, Е А

2016-01-25

Ceramicrete, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid-base reaction between magnesium oxide and mono potassium phosphate that has a struvite-K mineral structure. In this study, we demonstrate that this crystalline structure is ideal for incorporating radioactive Cs into a Ceramicrete matrix. This is accomplished by partially replacing K by Cs in the struvite-K structure, thus forming struvite-(K, Cs) mineral. X-ray diffraction and thermo-gravimetric analyses are used to confirm such a replacement. The resulting product is non-leachable and stable at high temperatures, and hence it is an ideal matrix for immobilizing Cs found in high-activity nuclear waste streams. The product can also be used for immobilizing secondary waste streams generated during glass vitrification of spent fuel, or the method described in this article can be used as a pretreatment method during glass vitrification of high level radioactive waste streams. Furthermore, it suggests a method of producing safe commercial radioactive Cs sources. Copyright © 2015 Elsevier B.V. All rights reserved.

Chemical immobilization of fission products reactive with nuclear reactor components

International Nuclear Information System (INIS)

Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

1975-01-01

This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

Change in blood glucose level in rats after immobilization

Science.gov (United States)

Platonov, R. D.; Baskakova, G. M.; Chepurnov, S. A.

1981-01-01

Experiments were carried out on male white rats divided into four groups. In group one the blood glucose level was determined immediately after immobilization. In the other three groups, two hours following immobilization, the blood glucose level was determined every 20 minutes for 3 hours 40 minutes by the glucose oxidase method. Preliminary immobilization for 2 hours removed the increase in the blood glucose caused by the stress reaction. By the 2nd hour of immobilization in the presence of continuing stress, the blood glucose level stabilized and varied within 42 + or - 5.5 and 47 + or - 8.1 mg %. Within 2 hours after the immobilization, the differences in the blood glucose level of the rats from the control groups were statistically insignificant.

Immobilization of high-level wastes into sintered glass: 2

International Nuclear Information System (INIS)

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

1987-01-01

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

Solidification of Savannah River Plant high level waste

International Nuclear Information System (INIS)

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

1981-11-01

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

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

International Nuclear Information System (INIS)

Cowan, S.P.

1985-01-01

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

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

International Nuclear Information System (INIS)

Cornman, W.R.

1980-06-01

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

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

International Nuclear Information System (INIS)

1988-01-01

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

Storage of High Level Nuclear Waste in Germany

Directory of Open Access Journals (Sweden)

Dietmar P. F. Möller

2007-01-01

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

Nuclear waste immobilization in iron phosphate glasses

International Nuclear Information System (INIS)

Garcia, D.A.; Rodriguez, Diego A.; Menghini, Jorge E.; Bevilacqua, Arturo

2007-01-01

Iron-phosphate glasses have become important in the nuclear waste immobilization area because they have some advantages over silicate-based glasses, such as a lower processing temperature and a higher nuclear waste load without losing chemical and mechanical properties. Structure and chemical properties of iron-phosphate glasses are determined in terms of the main components, in this case, phosphate oxide along with the other oxides that are added to improve some of the characteristics of the glasses. For example, Iron oxide improves chemical durability, lead oxide lowers fusion temperature and sodium oxide reduces viscosity at high temperature. In this work a study based on the composition-property relations was made. We used different techniques to characterize a series of iron-lead-phosphate glasses with uranium and aluminium oxide as simulated nuclear waste. We used the Arquimedes method to determine the bulk density, differential temperature analysis (DTA) to determine both glass transition temperature and crystallization temperature, dilatometric analysis to calculate the linear thermal expansion coefficient, chemical durability (MCC-1 test) and X-ray diffraction (XRD). We also applied some theoretic models to calculate activation energies associated with the glass transition temperature and crystallization processes. (author)

Immobilization needs and technology programs

International Nuclear Information System (INIS)

Gray, L.W.; Kan, T.; Shaw, H.; Armantrout, G.

1995-01-01

In the aftermath of the Cold War, the US and Russia agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, DOE has undertaken a multifaceted study to select options for storage and disposition of plutonium in keeping with US policy that plutonium must be subjected to the highest standards of safety, security, and accountability. One alternative being considered is immobilization. To arrive at a suitable immobilization form, we first reviewed published information on high-level waste immobilization technologies and identified 72 possible plutonium immobilization forms to be prescreened. Surviving forms were further screened using multi-attribute utility analysis to determine the most promising technology families. Promising immobilization families were further evaluated to identify chemical, engineering, environmental, safety, and health problems that remain to be solved prior to making technical decisions as to the viability of using the form for long- term disposition of plutonium. From this evaluation, a detailed research and development plan has been developed to provide answers to these remaining questions

Development of aluminosilicate and borosilicate glasses as matrices for CANDU high-level waste

International Nuclear Information System (INIS)

Strathdee, G.G.; McIntyre, N.S.; Taylor, P.

1979-01-01

This paper covers the results of analyses of two radioactive nepheline syenite glass blocks recovered from in-ground leaching experiments at the Chalk River Nuclear Laboratories. Current research on borosilicate glasses for immobilization of high-level waste is also described

Remote handling in the Plutonium Immobilization Project: Puck handling

International Nuclear Information System (INIS)

Brault, J.R.

2000-01-01

Since the break up of the Soviet Union at the end of the Cold War, the US and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

Preliminary evaluation of alternative forms for immobilization of Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Stone, J.A.; Goforth, S.T. Jr.; Smith, P.K.

1979-12-01

An evaluation of available information on eleven alternative solid forms for immobilization of SRP high-level waste has been completed. Based on the assessment of both product and process characteristics, four forms were selected for more detailed evaluation: (1) borosilicate glass made in the reference process, (2) a high-silica glass made from a porous glass matrix, (3) crystalline ceramics such as supercalcine or SYNROC, and (4) ceramics coated with an impervious barrier. The assessment includes a discussion of product and process characteristics for each of the eleven forms, a cross comparison of these characteristics for the forms, and the bases for selecting the most promising forms for further study

On immobilization of high-level waste in an Y–Al garnet-based cermet matrix in SHS conditions

OpenAIRE

Konovalov, E.E.; Lastov, A.I.; Nerozin, N.A.

2015-01-01

A method of high-level waste (HLW) radionuclide immobilization in a long-life matrix based on Y–Al garnet, a material highly chemically resistant to natural environments, has been developed for the ultimate HLW isolation from the environment. Model systems containing Ce, Nd, Sm, Zr, Mo, 238U, and 241Am were used in the study as simulators of HLW radionuclides. An energy-saving technology of self-propagating high-temperature synthesis (SHS) was employed to synthesize the matrix material with f...

Alternatives generation and analysis report for immobilized low-level waste interim storage architecture

Energy Technology Data Exchange (ETDEWEB)

Burbank, D.A., Westinghouse Hanford

1996-09-01

The Immobilized Low-Level Waste Interim Storage subproject will provide storage capacity for immobilized low-level waste product sold to the U.S. Department of Energy by the privatization contractor. This report describes alternative Immobilized Low-Level Waste storage system architectures, evaluation criteria, and evaluation results to support the Immobilized Low-Level Waste storage system architecture selection decision process.

Vitrification of low level and mixed (radioactive and hazardous) wastes: Lessons learned from high level waste vitrification

International Nuclear Information System (INIS)

Jantzen, C.M.

1994-01-01

Borosilicate glasses will be used in the USA and in Europe immobilize radioactive high level liquid wastes (HLLW) for ultimate geologic disposal. Simultaneously, tehnologies are being developed by the US Department of Energy's (DOE) Nuclear Facility sites to immobilize low-level and mixed (radioactive and hazardous) wastes (LLMW) in durable glass formulations for permanent disposal or long-term storage. Vitrification of LLMW achieves large volume reductions (86--97 %) which minimize the associated long-term storage costs. Vitrification of LLMW also ensures that mixed wastes are stabilized to the highest level reasonably possible, e.g. equivalent to HLLW, in order to meet both current and future regulatory waste disposal specifications The tehnologies being developed for vitrification of LLMW rely heavily on the technologies developed for HLLW and the lessons learned about process and product control

Conceptual process for immobilizing defense high level wastes in SYNROC-D

International Nuclear Information System (INIS)

Anon.

1981-01-01

It is believed that the immobilization of defense wastes in SYNROC-D possesses important advantages over an alternative process which involves immobilizing the sludges in borosilicate glass. (1) It is possible to immobilize about 3 times the weight of sludge in a given volume of SYNROC-D as compared to borosilicate glass. The costs of fabrications, transport and ultimate geologic storage are correspondingly reduced; (2) the mineral assemblage of SYNROC-D is vastly more stable in the presence of groundwaters than are borosilicate glasses. The long-lived actinide elements, in particular, are immobilized much more securely in SYNROC-D than in glass; and (3) SYNROC-D is composed of thermodynamically compatible phases which possess crystal structures identical to those of natural minerals which are known to have survived in geological environments at elevated pressures and temperatures for periods of 500 to 2000 million years and to have retained radioactive elements quantitatively for these periods despite strong radiation damage. It is this evidence, provided by nature herself, which can demonstrate to the community that the shorter times required for radwaste immobilization under the much less extreme pressure, temperature conditions present in a suitable geological repository can be successfully achieved. Glass, as a waste-form, is intrinsically incapable of providing this assurance

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

International Nuclear Information System (INIS)

CALMUS, R.B.

2000-01-01

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

Evaluation and selection of candidate high-level waste forms

International Nuclear Information System (INIS)

1982-03-01

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

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

International Nuclear Information System (INIS)

Vinjamuri, K.

1990-01-01

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

Design-Only Conceptual Design Report: Plutonium Immobilization Plant

International Nuclear Information System (INIS)

DiSabatino, A.; Loftus, D.

1999-01-01

This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

Disposal of high level nuclear wastes: Thermodynamic equilibrium and environment ethics

Institute of Scientific and Technical Information of China (English)

RANA Mukhtar Ahmed

2009-01-01

Contamination of soil, water or air, due to a failure of containment or disposal of high level nuclear wastes, can potentially cause serious hazards to the environment or human health. Essential elements of the environment and radioactivity dangers to it are illustrated. Issues of high level nuclear waste disposal are discussed with a focus on thermodynamic equilibrium and environment ethics. Major aspects of the issues are analyzed and described briefly to build a perception of risks involved and ethical implications. Nuclear waste containment repository should be as close as possible to thermodynamic equilibrium. A clear demonstration about safety aspects of nuclear waste management is required in gaining public and political confidence in any possible scheme of permanent disposal. Disposal of high level nuclear waste offers a spectrum of environment connected challenges and a long term future of nuclear power depends on the environment friendly solution of the problem of nuclear wastes.

Geology of high-level nuclear waste disposal: an introduction

International Nuclear Information System (INIS)

Roxbugh, I.S.

1987-01-01

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

Disposal of high level nuclear wastes: thermodynamic equilibrium and environment ethics

International Nuclear Information System (INIS)

Rana, M.A.

2009-01-01

Contamination of soil, water or air, due to a failure of containment or disposal of high level nuclear wastes, can potentially cause serious hazards to the environment or human health. Essential elements of the environment and radioactivity dangers to it are illustrated. Issues of high level nuclear waste disposal are discussed with a focus on thermodynamic equilibrium and environment ethics. Major aspects of the issues are analyzed and described briefly to build a perception of risks involved and ethical implications. Nuclear waste containment repository should be as close as possible to thermodynamic equilibrium. A clear demonstration about safety aspects of nuclear waste management is required in gaining public and political confidence in any possible scheme of permanent disposal. Disposal of high level nuclear waste offers a spectrum of environment connected challenges and a long term future of nuclear power depends on the environment friendly solution of the problem of nuclear wastes. (authors)

Experience in industrial operation of the plant for immobilizing radioactive wastes in thermosetting resins at the Ardennes Nuclear Power Station

International Nuclear Information System (INIS)

Haller, P.; Romestain, P.; Bruant, J.P.

1983-01-01

The French Atomic Energy Commission (CEA) has developed, at the Grenoble Centre for Nuclear Studies, a procedure for immobilizing low- and intermediate-level wastes in thermosetting resins of the polyester or epoxy types. To demonstrate feasibility on an industrial scale, a pilot plant has been set up at the effluent treatment station of the Ardennes Franco-Belgium Nuclear Power Station (SENA), which is a 305 MW(e) PWR type. Assembly work began in January 1979. After a period devoted to final adjustments and operation with inactive products, conditioning of active products began in January 1981. In the paper, the methods of conditioning the three types of waste (evaporation concentrates, ion exchange resins and filter cartridges) are described, experience of the start-up and operation of the plant is reported and the principal results of coating characterization tests are given. The results of tests on active and inactive products show that the characteristics of the materials obtained on an industrial scale match those of laboratory products and confirm their high quality with regard to mechanical behaviour, fire resistance, homogeneity and low-leachability. Industrial experience and economic comparisons show that the process of immobilizing waste from nuclear power stations in thermosetting resins offers an extremely interesting alternative to classical methods of conditioning. (author)

Effect of sulfonylureas administered centrally on the blood glucose level in immobilization stress model.

Science.gov (United States)

Sharma, Naveen; Sim, Yun-Beom; Park, Soo-Hyun; Lim, Su-Min; Kim, Sung-Su; Jung, Jun-Sub; Hong, Jae-Seung; Suh, Hong-Won

2015-05-01

Sulfonylureas are widely used as an antidiabetic drug. In the present study, the effects of sulfonylurea administered supraspinally on immobilization stress-induced blood glucose level were studied in ICR mice. Mice were once enforced into immobilization stress for 30 min and returned to the cage. The blood glucose level was measured 30, 60, and 120 min after immobilization stress initiation. We found that intracerebroventricular (i.c.v.) injection with 30 µg of glyburide, glipizide, glimepiride or tolazamide attenuated the increased blood glucose level induced by immobilization stress. Immobilization stress causes an elevation of the blood corticosterone and insulin levels. Sulfonylureas pretreated i.c.v. caused a further elevation of the blood corticosterone level when mice were forced into the stress. In addition, sulfonylureas pretreated i.c.v. alone caused an elevation of the plasma insulin level. Furthermore, immobilization stress-induced insulin level was reduced by i.c.v. pretreated sulfonylureas. Our results suggest that lowering effect of sulfonylureas administered supraspinally against immobilization stress-induced increase of the blood glucose level appears to be primarily mediated via elevation of the plasma insulin level.

SYNROC process. A geochemical approach to nuclear waste immobilization

Energy Technology Data Exchange (ETDEWEB)

Ringwood, A E; Kesson, S E; Ware, N G; Hibberson, W O; Major, A [Australian National Univ., Canberra. Research School of Earth Sciences

1979-08-01

The SYNROC process is proposed to immobilize high-level wastes as dilute solid solutions in the constituent minerals of a synthetic rock formed from a mixture of oxides. New modification of the SYNROC was developed. Experiments showed that the entire spectra of high-level waste elements can be incorporated in the crystal lattices of Ba-hollandite, perovskite and zirconolite. This titanate assemblage has been proved to be exceptionally resistant to hydrothermal leaching, and in this respect, amongst others, it is demonstrably superior to alternative ceramic waste forms and to borosilicate glasses. The relative stability of various waste forms was compared in hydrothermal leaching experiments using both pure water and 10 w/o NaCl solution. Borosilicate glasses were almost completely decomposed and disintegrated after only 24 hours at 350 deg C and 1000 bars, and the extensive loss of hazardous high-level waste elements occurred. The phase pollucite in ceramic waste forms began to decompose at 400 deg C. The hollandite-perovskite-zirconolite SYNROC assemblage was proved to be exceptionally resistant to leaching, surviving invariably the extreme conditions up to 900 deg C and 5000 bars. Geochemical studies of the naturally-occurring minerals containing radwaste elements are relevant to the problem of radiation damage to SYNROC phases. These imply that the 2-particle flux in SYNROC is unlikely to be enough to impair the ability to immobilize radwaste for the required period. The production of SYNROC is explained in detail. The SYNROC phases have the structures analogous to the natural minerals which have survived a variety of geological conditions for millions of years while retaining certain high-level waste elements in their crystal lattices.

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

International Nuclear Information System (INIS)

Calmus, R.B.

1998-01-01

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

Experimental study on cesium immobilization in struvite structures

International Nuclear Information System (INIS)

Wagh, Arun S.; Sayenko, S.Y.; Shkuropatenko, V.A.; Tarasov, R.V.; Dykiy, M.P.; Svitlychniy, Y.O.; Virych, V.D.; Ulybkina, E.A.

2016-01-01

Graphical abstract: X-ray diffraction patterns of Ceramicrete forms, green representing struvite-K, and red, struvite-(K,Cs) with 10 wt.% CsCl in it. Cs substitutes partially for K, which immobilizes Cs at room temperature by the acid–base reaction. - Highlights: • Struvite structure of Ceramicrete is an excellent host of radioactive cesium. • The volatility problem of cesium can be avoided by this method. • This method can be used to produce cesium waste forms in ambient conditions. • It can also be used to pretreat cesium in glass vitrification technology. • It also provides a method to produce safe sealed radioactive sources of cesium. - Abstract: Ceramicrete, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid–base reaction between magnesium oxide and mono potassium phosphate that has a struvite-K mineral structure. In this study, we demonstrate that this crystalline structure is ideal for incorporating radioactive Cs into a Ceramicrete matrix. This is accomplished by partially replacing K by Cs in the struvite-K structure, thus forming struvite-(K, Cs) mineral. X-ray diffraction and thermo-gravimetric analyses are used to confirm such a replacement. The resulting product is non-leachable and stable at high temperatures, and hence it is an ideal matrix for immobilizing Cs found in high-activity nuclear waste streams. The product can also be used for immobilizing secondary waste streams generated during glass vitrification of spent fuel, or the method described in this article can be used as a pretreatment method during glass vitrification of high level radioactive waste streams. Furthermore, it suggests a method of producing safe commercial radioactive Cs sources.

Experimental study on cesium immobilization in struvite structures

Energy Technology Data Exchange (ETDEWEB)

Wagh, Arun S., E-mail: asw@anl.gov [Environmental Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, IL 60439 (United States); Sayenko, S.Y.; Shkuropatenko, V.A.; Tarasov, R.V.; Dykiy, M.P.; Svitlychniy, Y.O.; Virych, V.D.; Ulybkina, E.A. [National Science Center, Kharkov Institute of Physics and Technology, Kharkov (Ukraine)

2016-01-25

Graphical abstract: X-ray diffraction patterns of Ceramicrete forms, green representing struvite-K, and red, struvite-(K,Cs) with 10 wt.% CsCl in it. Cs substitutes partially for K, which immobilizes Cs at room temperature by the acid–base reaction. - Highlights: • Struvite structure of Ceramicrete is an excellent host of radioactive cesium. • The volatility problem of cesium can be avoided by this method. • This method can be used to produce cesium waste forms in ambient conditions. • It can also be used to pretreat cesium in glass vitrification technology. • It also provides a method to produce safe sealed radioactive sources of cesium. - Abstract: Ceramicrete, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid–base reaction between magnesium oxide and mono potassium phosphate that has a struvite-K mineral structure. In this study, we demonstrate that this crystalline structure is ideal for incorporating radioactive Cs into a Ceramicrete matrix. This is accomplished by partially replacing K by Cs in the struvite-K structure, thus forming struvite-(K, Cs) mineral. X-ray diffraction and thermo-gravimetric analyses are used to confirm such a replacement. The resulting product is non-leachable and stable at high temperatures, and hence it is an ideal matrix for immobilizing Cs found in high-activity nuclear waste streams. The product can also be used for immobilizing secondary waste streams generated during glass vitrification of spent fuel, or the method described in this article can be used as a pretreatment method during glass vitrification of high level radioactive waste streams. Furthermore, it suggests a method of producing safe commercial radioactive Cs sources.

High-level nuclear waste disposal

International Nuclear Information System (INIS)

Burkholder, H.C.

1985-01-01

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

Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor

International Nuclear Information System (INIS)

Garcia Martinez, H.

1999-01-01

In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 L ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor

International Nuclear Information System (INIS)

Garcia M, H.; Emeterio H, M.; Canizal S, C.

1999-01-01

This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

Handling and storage of conditioned high-level wastes

International Nuclear Information System (INIS)

1983-01-01

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

RAD18 and associated proteins are immobilized in nuclear foci in human cells entering S-phase with ultraviolet light-induced damage

International Nuclear Information System (INIS)

Watson, Nicholas B.; Nelson, Eric; Digman, Michelle; Thornburg, Joshua A.; Alphenaar, Bruce W.; McGregor, W. Glenn

2008-01-01

Proteins required for translesion DNA synthesis localize in nuclear foci of cells with replication-blocking lesions. The dynamics of this process were examined in human cells with fluorescence-based biophysical techniques. Photobleaching recovery and raster image correlation spectroscopy experiments indicated that involvement in the nuclear foci reduced the movement of RAD18 from diffusion-controlled to virtual immobility. Examination of the mobility of REV1 indicated that it is similarly immobilized when it is observed in nuclear foci. Reducing the level of RAD18 greatly reduced the focal accumulation of REV1 and reduced UV mutagenesis to background frequencies. Fluorescence lifetime measurements indicated that RAD18 and RAD6A or polη only transferred resonance energy when these proteins colocalized in damage-induced nuclear foci, indicating a close physical association only within such foci. Our data support a model in which RAD18 within damage-induced nuclear foci is immobilized and is required for recruitment of Y-family DNA polymerases and subsequent mutagenesis. In the absence of damage these proteins are not physically associated within the nucleoplasm

Managing the high level waste nuclear regulatory commission licensing process

International Nuclear Information System (INIS)

Baskin, K.P.

1992-01-01

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

Answers to your questions on high-level nuclear waste

International Nuclear Information System (INIS)

1987-11-01

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

Waste incineration and immobilization for nuclear facilities, April--September 1977

International Nuclear Information System (INIS)

Johnson, A.J.; Fong, L.Q.

1978-01-01

Fluidized bed incineration and waste immobilization processes are being developed to process the types of waste expected from nuclear facilities. An air classification system has been developed to separate tramp metal from shredded combustible solid waste prior to the waste being fed to a fluidized-bed pilot-plant incinerator. Used organic ion exchange resin with up to 55 percent water has been effectively burned in the fluidized bed incinerator. Various methods of feeding waste into the incinerator were investigated as alternatives to the present compression screw; an extrusion ram was found to suffer extensive damage from hard particles in tested waste. A bench-scale continuous waste immobilization process has been operated and has produced glass from incinerator residue and other types of waste materials

Remote material handling in the Plutonium Immobilization Project. Revision 1

International Nuclear Information System (INIS)

Brault, J.R.

2000-01-01

With the downsizing of the US and Russian nuclear stockpiles, large quantities of weapons-usable plutonium in the US are being declared excess and will be disposed of by the Department of Energy Fissile Materials Disposition Program. To implement this program, DOE has selected the Savannah River Site (SRS) for the construction and operation of three new facilities: pit disassembly and conversion; mixed oxide fuel fabrication; and plutonium immobilization. The Plutonium Immobilization Project (PIP) will immobilize a portion of the excess plutonium in a hybrid ceramic and glass form containing high level waste for eventual disposal in a geologic repository. The PIP is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. Processing technology for the PIP is being developed jointly by the Lawrence Livermore National Laboratory and Westinghouse Savannah River Company. This paper will discuss development of the automated unpacking and sorting operations in the conversion area, and the automated puck and tray handling operations in the first stage immobilization area. Due to the high radiation levels and toxicity of the materials to be disposed of, the PIP will utilize automated equipment in a contained (glovebox) facility. Most operations involving plutonium-bearing materials will be performed remotely, separating personnel from the radiation source. Source term materials will be removed from the operations during maintenance. Maintenance will then be performed hands on within the containment using glove ports

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Radiation Effects in Nuclear Waste Materials

International Nuclear Information System (INIS)

Weber, William J.; Wang, Lumin; Hess, Nancy J.; Icenhower, Jonathan P.; Thevuthasan, Suntharampillai

2003-01-01

The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials

Radiation Effects in Nuclear Waste Materials

International Nuclear Information System (INIS)

Weber, William J.

2005-01-01

The objective of this project is to develop a fundamental understanding of radiation effects in glasses and ceramics, as well as the influence of solid-state radiation effects on aqueous dissolution kinetics, which may impact the performance of nuclear waste forms and stabilized nuclear materials. This work provides the underpinning science to develop improved glass and ceramic waste forms for the immobilization and disposition of high-level tank waste, excess plutonium, plutonium residues and scrap, other actinides, and other nuclear waste streams. Furthermore, this work is developing develop predictive models for the performance of nuclear waste forms and stabilized nuclear materials. Thus, the research performed under this project has significant implications for the immobilization of High-Level Waste (HLW) and Nuclear Materials, two mission areas within the Office of Environmental Management (EM). With regard to the HLW mission, this research will lead to improved understanding of radiation-induced degradation mechanisms and their effects on dissolution kinetics, as well as development of predictive models for waste form performance. In the Nuclear Materials mission, this research will lead to improvements in the understanding of radiation effects on the chemical and structural properties of materials for the stabilization and long-term storage of plutonium, highly-enriched uranium, and other actinides. The research uses plutonium incorporation, ion-beam irradiation, and electron-beam irradiation to simulate the effects of alpha decay and beta decay on relevant glasses and ceramics. The research under this project has the potential to result in improved glass and ceramic materials for the stabilization and immobilization of high-level tank waste, plutonium residues and scraps, surplus weapons plutonium, highly-enriched uranium, other actinides, and other radioactive materials

Ceramic Single Phase High-Level Nuclear Waste Forms: Hollandite, Perovskite, and Pyrochlore

Science.gov (United States)

Vetter, M.; Wang, J.

2017-12-01

The lack of viable options for the safe, reliable, and long-term storage of nuclear waste is one of the primary roadblocks of nuclear energy's sustainable future. The method being researched is the incorporation and immobilization of harmful radionuclides (Cs, Sr, Actinides, and Lanthanides) into the structure of glasses and ceramics. Borosilicate glasses are the main waste form that is accepted and used by today's nuclear industry, but they aren't the most efficient in terms of waste loading, and durability is still not fully understood. Synroc-phase ceramics (i.e. hollandite, perovskite, pyrochlore, zirconolite) have many attractive qualities that glass waste forms do not: high waste loading, moderate thermal expansion and conductivity, high chemical durability, and high radiation stability. The only downside to ceramics is that they are more complex to process than glass. New compositions can be discovered by using an Artificial Neural Network (ANN) to have more options to optimize the composition, loading for performance by analyzing the non-linear relationships between ionic radii, electronegativity, channel size, and a mineral's ability to incorporate radionuclides into its structure. Cesium can be incorporated into hollandite's A-site, while pyrochlore and perovskite can incorporate actinides and lanthanides into their A-site. The ANN is used to predict new compositions based on hollandite's channel size, as well as the A-O bond distances of pyrochlore and perovskite, and determine which ions can be incorporated. These new compositions will provide more options for more experiments to potentially improve chemical and thermodynamic properties, as well as increased waste loading capabilities.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Barium borosilicate glass - a potential matrix for immobilization of sulfate bearing high-level radioactive liquid waste

International Nuclear Information System (INIS)

Kaushik, C.P.; Mishra, R.K.; Sengupta, P.; Kumar, Amar; Das, D.; Kale, G.B.; Raj, Kanwar

2006-01-01

Borosilicate glass formulations adopted worldwide for immobilization of high-level radioactive liquid waste (HLW) is not suitable for sulphate bearing HLW, because of its low solubility in such glass. A suitable glass matrix based on barium borosilicate has been developed for immobilization of sulphate bearing HLW. Various compositions based on different glass formulations were made to examine compatibility with waste oxide with around 10 wt% sulfate content. The vitrified waste product obtained from barium borosilicate glass matrix was extensively evaluated for its characteristic properties like homogeneity, chemical durability, glass transition temperature, thermal conductivity, impact strength, etc. using appropriate techniques. Process parameters like melt viscosity and pour temperature were also determined. It is found that SB-44 glass composition (SiO 2 : 30.5 wt%, B 2 O 3 : 20.0 wt%, Na 2 O: 9.5 wt% and BaO: 19.0 wt%) can be safely loaded with 21 wt% waste oxide without any phase separation. The other product qualities of SB-44 waste glass are also found to be on a par with internationally adopted waste glass matrices. This formulation has been successfully implemented in plant scale

Immobilization in cement of ion exchange resins from Spanish nuclear reactors

International Nuclear Information System (INIS)

Huebra, A.G. de la; Murillo, R.; Ortiz, S.J.

1990-01-01

Ion exchange materials used at nuclear power plants can be immobilized in cements less expensive than polymer matrices. Cement solidification of spent ion exchange resins shows swelling and cracking troubles (during setting time, or of storage). The objective of this study was to select the types of cement that produce the best quality on immobilization of three kinds of resins and to set up cement formulations containing the maximum possible loading of resin. Four cements were selected to carried out the study. After a study of hydration-dehydration phenomena of ion exchange resins, a systematic work has been carried out on immobilization. Tests were performed to study compressive strength and underwater stability by changing water/cement ratio and resin/cement ratio. Mixtures made with water, cement and resin only were loaded with 10% by weight dry resin. Mixtures with higher loadings show poor workability. Tests were carried out by adding organic plasticizers and silica products to improve waste loading. Plasticizers reduced water demand and silica products permit the use of more water. Leaching tests have been performed at 40 O C. In conclusion Blast Furnace Slag is the best cement for immobilization of ion exchange resin both bead and powdered form for mechanical strength, stability and leaching

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

International Nuclear Information System (INIS)

Cornman, W.R.

1981-12-01

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

Engineering-scale vitrification of commercial high-level waste

International Nuclear Information System (INIS)

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

1980-04-01

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

High-level nuclear waste disposal: Ethical considerations

International Nuclear Information System (INIS)

Maxey, M.N.

1985-01-01

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

Immobilisation of high level nuclear reactor wastes in SYNROC

Energy Technology Data Exchange (ETDEWEB)

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

1979-03-15

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

Management of the high-level nuclear power facilities

International Nuclear Information System (INIS)

Preda, Marin

2003-05-01

This thesis approaches current issues in the management of the high power nuclear facilities and as such it appears to be important particularly for nuclear power plant operation topics. Of special interest are the failure events entailing possible catastrophic situations. The contents is structured onto ten chapters. The first chapter describes the operation regimes of the nuclear high power facilities. Highlighted here are the thesis scope and the original features of the work. The second chapter deals with operational policies developed in order to ensure the preventive maintenance of the nuclear installations. Also managing structures are described devoted to practical warranting the equipment safety function of non-classical power stations. In the third chapter cases of nuclear accidents are analyzed especially stressing the probabilistic risk and the operation regimes having in view the elimination of catastrophic events. In the fourth and fifth chapters the control of nuclear radiation emission is treated focusing the quality issue of nuclear installations required to avoid hazardous effects at level of nuclear reactor operation stage. At the same time set of operational measures is given here for preventing risks, catastrophes and chaotic situations. The chapter five presents both theoretical and practical approaches of the nuclear reactor core management concerning particularly the fuel testing, the water primary system and the quality of the involved equipment. In the sixth and seventh chapters issues of risk-quality correlations are approached as well as the structure of expert systems for monitoring the operational regimes of nuclear facilities. The efficiency of the power systems with nuclear injection is discussed and some original ideas developed in this work are evidenced in the eighth and ninth chapters. Presented are here both the operational principles and models of raising the efficiency of the interconnected nuclear stations and prices' policy

Immobilization as a route to surplus fissile materials disposition

International Nuclear Information System (INIS)

Gray, L.W.; Kan, T.

1995-01-01

In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, DOE has undertaken a multifaceted study to select options for storage and disposition of plutonium (Pu) in keeping with the national policy that Pu must be subjected to the highest standards of safety, security, and accountability. One alternative being considered is immobilization. To arrive at a suitable immobilization form, the authors first reviewed published information on high-level waste (HLW) immobilization technologies in order to identify 72 possible Pu immobilization forms to be prescreened. Surviving forms were screened using multiattribute analysis to determine the most promising technologies. Promising immobilization families were further evaluated to identify chemical, engineering, environmental, safety, and health problems that remain to be solved prior to making technical decisions as to the viability of using the form for long-term disposition of plutonium. All data, analyses, and reports are being provided to the DOE Fissile Materials Disposition Project Office to support the Record of Decision that is anticipated in the fourth quarter of FY96

Immobilization of high level nuclear reactor wastes in SYNROC: a current appraisal

International Nuclear Information System (INIS)

Oversby, V.M.; Ringwood, A.E.

1981-01-01

Results are presented for leach testing at 95 0 C and 200 0 C of SYNROC containing 9% and 20% simulated high level radioactive waste, synthetic hollandite and pervoskite samples, and natural zirconolite and pervoskite samples. Single phase synthetic minerals show much higher leach rates than natural mineral samples and polyphase SYNROC samples. Natural zirconolite samples with low radiation damage have leach rates at 200 0 C based on U which are identical to those measured on SYNROC samples. Natural zirconolites with very large accumulated α dose and radiation damage have leach rates at 200 0 C which are only 5 times higher than those of low dose samples

Statement of work for the immobilized high-level waste transportation system, Project W-464

Energy Technology Data Exchange (ETDEWEB)

Mouette, P.

1998-06-24

The objective of this Statement of Work (SOW) is to present the scope, the deliverables, the organization, the technical and schedule expectations for the development of a Package Design Criteria (PDC), cost and schedule estimate for the acquisition of a transportation system for the Immobilized High-Level Waste (IHLW). This transportation system which includes the truck, the trailer, and a shielded cask will be used for on-site transportation of the IHLW canisters from the private vendor vitrification facility to the Hanford Site interim storage facility, i.e., vaults 2 and 3 of the Canister Storage Building (CSB). This Statement of Work asks Waste Management Federal Services, Inc., Northwest Operations, to provide Project W-464 with a Design Criteria Document, plus a life-cycle schedule and cost estimate for the acquisition of a transportation system (shielded cask, truck, trailer) for IHLW on-site transportation.

Single Phase Melt Processed Powellite (Ba,Ca) MoO{sub 4} For The Immobilization Of Mo-Rich Nuclear Waste

Energy Technology Data Exchange (ETDEWEB)

Brinkman, Kyle [Savannah River Site (SRS), Aiken, SC (United States); Marra, James [Savannah River Site (SRS), Aiken, SC (United States); Fox, Kevin [Savannah River Site (SRS), Aiken, SC (United States); Reppert, Jason [Savannah River Site (SRS), Aiken, SC (United States); Crum, Jarrod [Paci fic Northwest National Laboratory , Richland, WA (United States); Tang, Ming [Los Alamos National Laboratory , Los Alamos, NM (United States)

2012-09-17

Crystalline and glass composite materials are currently being investigated for the immobilization of combined High Level Waste (HLW) streams resulting from potential commercial fuel reprocessing scenarios. Several of these potential waste streams contain elevated levels of transition metal elements such as molybdenum (Mo). Molybdenum has limited solubility in typical silicate glasses used for nuclear waste immobilization. Under certain chemical and controlled cooling conditions, a powellite (Ba,Ca)MoO{sub 4} crystalline structure can be formed by reaction with alkaline earth elements. In this study, single phase BaMoO{sub 4} and CaMoO{sub 4} were formed from carbonate and oxide precursors demonstrating the viability of Mo incorporation into glass, crystalline or glass composite materials by a melt and crystallization process. X-ray diffraction, photoluminescence, and Raman spectroscopy indicated a long range ordered crystalline structure. In-situ electron irradiation studies indicated that both CaMoO{sub 4} and BaMoO{sub 4} powellite phases exhibit radiation stability up to 1000 years at anticipated doses with a crystalline to amorphous transition observed after 1 X 10{sup 13} Gy. Aqueous durability determined from product consistency tests (PCT) showed low normalized release rates for Ba, Ca, and Mo (<0.05 g/m{sup 2}).

CHARACTER OF THE CHANGES IN FEAR MOTIVATED DECLARATIVE MEMORY IN THE HIGH IMMOBILIZATION "DEPRESSIVE" RATS.

Science.gov (United States)

Nachkebia, N; Shavgulidze, M; Babilodze, M; Chkhartishvili, E; Rogava, N

2016-10-01

Present study investigated possible differences in the learning and memory of declarative memory task in rats selected according to the differences in immobilization response that is in high immobilization "depressive" and low immobilization "non-depressive" rats. Understanding the character of learning and memory disturbances in basal conditions of animal models of depression is still very topical for more intimate definition of the pathophysiology of major depressive disorder and appropriate searching the ways of its correction. Experiments were carried out on the adult white wild rats (with the weight 200-250 g, n=20). Selection of rats according to the level of immobilization was made by means of forced swim test. Learning and memory disturbances were studied using passive avoidance test that is fear motivated one trial declarative memory task. It was shown by us that 100% of low immobilization "non-depressive" rats remember painful stimulation and therefore they are not enter in the dark compartment during whole period of observation during testing session. Behavior of high immobilization "depressive" rats is not similar in passive avoidance camera; 50% of "depressive" rats, with long escape latency during training session (92±10 sec), remember painful stimulation during testing session and therefore they are not enter in the dark compartment during whole observation period. The remaining 50%, that are not differ significantly from the low immobility "non-depressive" rats by the latency of escape (5±1 sec) during training session, are not able to remember painful stimulation during testing session and therefore they enter in the dark compartment with shortest escape latency (6±1 sec). In conclusion, high immobility "depressive" rats perform passive avoidance declarative memory task at the chance level that is a direct indicator for the serious disturbances of declarative memory mechanisms in "depressive" rats selected in forced swim test according to the

Nuclear level density

International Nuclear Information System (INIS)

Cardoso Junior, J.L.

1982-10-01

Experimental data show that the number of nuclear states increases rapidly with increasing excitation energy. The properties of highly excited nuclei are important for many nuclear reactions, mainly those that go via processes of the compound nucleus type. In this case, it is sufficient to know the statistical properties of the nuclear levels. First of them is the function of nuclear levels density. Several theoretical models which describe the level density are presented. The statistical mechanics and a quantum mechanics formalisms as well as semi-empirical results are analysed and discussed. (Author) [pt

76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

Science.gov (United States)

2011-06-16

... High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Public meeting... Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,'' and 73... Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage facilities. The draft regulatory...

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Development and evaluation of candidate high-level waste forms

International Nuclear Information System (INIS)

Bernadzikowski, T.A.

1981-01-01

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

Nuclear waste disposal: perspective of a geochemist

International Nuclear Information System (INIS)

Sengupta, Pranesh; Dey, G.K.

2011-01-01

Satisfying the growing requirement in an environment friendly way is one of the most important tasks we need to accomplish these days. Considering the restricted non-renewable energy resources and limited technological progresses achieved in the renewable energy sectors in India, nuclear energy appears to be one of the most lucrative solutions towards the forthcoming energy crisis. Successful implementation of nuclear energy program however requires careful execution of high level nuclear waste management activities. One very important aspect of this process is to identify and develop suitable inert matrix(ces) for conditioning of nuclear waste(s) using natural analogue studies. And this establishes the very vital linkage between geochemical studies and nuclear waste immobilization. One good example of such an interdisciplinary approach can be seen in the methodologies adopted for immobilization of sulfate bearing high level nuclear wastes (SO 4 -HLW). It has been reported on several occasions that sulfur-rich melt get separated from silicate melt within magma chamber. Similar process has also been witnessed within vitrification furnaces whenever an attempt has been made to condition SO 4 -HLW within borosilicate glass matrices. Since such liquid-liquid phase separation leads to multiple difficulties in connection to radionuclide immobilization and plant scale vitrification processes, solutions were sought from natural analogue studies. Such as integrated approach ultimately resulted in establishing two different methodologies e.g. (i) modifying the borosilicate network through introduction of Ba 2+ cation; a process being followed in India and (ii) using phosphatic melt as a host instead of borosilicate melt; a process being followed in Russia. Detail of these two routes and the geochemical linkage in nuclear waste immobilization will be discussed.(author)

Final disposal of high levels waste and spent nuclear fuel

International Nuclear Information System (INIS)

Gelin, R.

1984-05-01

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

Immobilization and Waste Form Product Acceptance for Low Level and TRU Waste Forms

International Nuclear Information System (INIS)

Holtzscheiter, E.W.; Harbour, J.R.

1998-05-01

The Tanks Focus Area is supporting technology development in immobilization of both High Level (HLW) and Low Level (LLW) radioactive wastes. The HLW process development at Hanford and Idaho is patterned closely after that of the Savannah River (Defense Waste Processing Facility) and West Valley Sites (West Valley Demonstration Project). However, the development and options open to addressing Low Level Waste are diverse and often site specific. To start, it is important to understand the breadth of Low Level Wastes categories

Novel immobilization process of a thermophilic catalase: efficient purification by heat treatment and subsequent immobilization at high temperature.

Science.gov (United States)

Xu, Juan; Luo, Hui; López, Claudia; Xiao, Jing; Chang, Yanhong

2015-10-01

The main goal of the present work is to investigate a novel process of purification and immobilization of a thermophilic catalase at high temperatures. The catalase, originated from Bacillus sp., was overexpressed in a recombinant Escherichia coli BL21(DE3)/pET28-CATHis and efficiently purified by heat treatment, achieving a threefold purification. The purified catalase was then immobilized onto an epoxy support at different temperatures (25, 40, and 55 °C). The immobilizate obtained at higher temperatures reached its maximum activity in a shorter time than that obtained at lower temperatures. Furthermore, immobilization at higher temperatures required a lower ionic strength than immobilization at lower temperatures. The characteristics of immobilized enzymes prepared at different temperatures were investigated. The high-temperature immobilizate (55 °C) showed the highest thermal stability, followed by the 40 °C immobilizate. And the high-temperature immobilizate (55 °C) had slightly higher operational stability than the 25 °C immobilizate. All of the immobilized catalase preparations showed higher stability than the free enzyme at alkaline pH 10.0, while the alkali resistance of the 25 °C immobilizate was slightly better than that of the 40 and 55 °C immobilizates.

Status of the high-level nuclear waste disposal program in Japan

International Nuclear Information System (INIS)

Uematsu, K.

1985-01-01

The Japan Atomic Energy Commission (JAEC) initiated a high-level radioactive waste disposal program in 1976. Since then, the Advisory Committee on Radioactive Waste Management of JAEC has revised the program twice. The latest revision was issued in 1984. The committee recommended a four-phase program and the last phase calls for the beginning of emplacement of the high-level nuclear waste into a selected repository in the Year 2000. The first phase is already completed, and the second phase of this decade calls for the selection of a candidate disposal site and the conducting of the RandD of waste disposal in an underground research laboratory and in a hot test facility. This paper covers the current status of the high-level nuclear waste disposal program in Japan

What are Spent Nuclear Fuel and High-Level Radioactive Waste?

International Nuclear Information System (INIS)

2002-01-01

Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository

High Hydrostatic Pressure Extract of Ginger Exerts Antistress Effects in Immobilization-Stressed Rats.

Science.gov (United States)

Moon, Sohee; Lee, Mak-Soon; Jung, Sunyoon; Kang, Bori; Kim, Seog-Young; Park, Seonyoung; Son, Hye-Yeon; Kim, Chong-Tai; Jo, Young-Hee; Kim, In-Hwan; Kim, Young Soon; Kim, Yangha

2017-09-01

Stress contributes to physiological changes such as weight loss and hormonal imbalances. The aim of the present study was to investigate antistress effects of high hydrostatic pressure extract of ginger (HPG) in immobilization-stressed rats. Male Sprague-Dawley rats (n = 24) were divided into three groups as follows: control (C), immobilization stress (2 h daily, for 2 weeks) (S), and immobilization stress (2 h daily, for 2 weeks) plus oral administration of HPG (150 mg/kg body weight/day) (S+G). Immobilization stress reduced the body weight gain and thymus weight by 50.2% and 31.3%, respectively, compared to the control group. The levels of serum aspartate transaminase, alanine transaminase, and corticosterone were significantly higher in the stress group, compared to the control group. Moreover, immobilization stress elevated the mRNA levels of tyrosine hydroxylase (Th), dopamine beta-hydroxylase (Dbh), and cytochrome P450 side-chain cleavage (P450scc), which are related to catecholamine and corticosterone synthesis in the adrenal gland. HPG administration also increased the body weight gain and thymus weight by 12.7% and 16.6%, respectively, compared to the stress group. Furthermore, the mRNA levels of Th, Dbh, phenylethanolamine-N-methyltransferase, and P450scc were elevated by the HPG treatment when compared to the stress group. These results suggest that HPG would have antistress effects partially via the reversal of stress-induced physiological changes and suppression of mRNA expression of genes related to corticosterone and catecholamine synthetic enzymes.

Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

Science.gov (United States)

Duffó, Gustavo S.; Farina, Silvia B.; Schulz, Fátima M.; Marotta, Francesca

2010-10-01

Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

Corrosion susceptibility of steel drums containing cemented intermediate level nuclear wastes

International Nuclear Information System (INIS)

Duffo, Gustavo S.; Farina, Silvia B.; Schulz, Fatima M.; Marotta, Francesca

2010-01-01

Cementation processes are used as immobilization techniques for low or intermediate level radioactive waste for economical and safety reasons and for being a simple operation. In particular, ion-exchange resins commonly used for purification of radioactive liquid waste from nuclear reactors are immobilized before being stored to improve the leach resistance of the waste matrix and to maintain mechanical stability. Combustible solid radioactive waste can be incinerated and the resulting ashes can also be immobilized before storage. The immobilized resins and ashes are then contained in steel drums that may undergo corrosion depending on the presence of certain contaminants. The work described in this paper was aimed at evaluating the corrosion susceptibility of steel drums in contact with cemented ion-exchange resins and incineration ashes containing different concentrations of aggressive species (mostly chloride and sulphate ions). A special type of specimen was designed to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix were monitored over a time period of 1 year. The results show the deleterious effect of chloride on the expected lifespan of the waste containers.

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

International Nuclear Information System (INIS)

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

1989-01-01

This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

International Nuclear Information System (INIS)

Peeler, D.

1999-01-01

'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

A survey of methods to immobilize tritium and carbon-14 arising from a nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Taylor, P.

1991-02-01

This report reviews the literature on methods to separate and immobilize tritium ( 3 H) and carbon-14 ( 14 C) released from U0 2 fuel in a nuclear fuel reprocessing plant. It was prepared as part of a broader review of fuel reprocessing waste management methods that might find future application in Canada. The calculated inventories of both 3 H and 14 C in used fuel are low; special measures to limit releases of these radionuclides from reprocessing plants are not currently in place, and may not be necessary in future. If required, however, several possible approaches to the concentration and immobilization of both radionuclides are available for development. Technology to control these radionuclides in reactor process streams is in general more highly developed than for reprocessing plant effluent, and some control methods may be adaptable to reprocessing applications

Investigation of waste form materials suitable for immobilizing actinide elements in high-level waste

International Nuclear Information System (INIS)

Hayakawa, Issei; Kamizono, Hiroshi

1992-07-01

The microstructure of waste form materials suitable for immobilizing actinide elements can be classified into the following two categories. (1) Actinide elements are immobilized in an crystallized matrix after the formation of solid solution or compounds. (2) Actinide elements are immobilized in a durable material by encapsulation. Based on crystal chemistry, durability data, phase diagrams, compositions of natural minerals, eleven oxide compounds and one non-oxide compound are pointed out to be new candidates included in category (1). The other survey on material compositions, manufacturing conditions and feasibility shows that SiC, glassy carbon, ZrO 2 , Ti-O-Si-C ceramics are preferable matrix materials included in category (2). Polymers and fine powders are suitable as starting materials for the encapsulation of actinide elements because of their excellent sinterability. (author) 50 refs

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

Science.gov (United States)

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

1984-04-11

Disclosed are lead-iron phosphate glasses containing a high level of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

Future health physics prospects in high-level nuclear waste management

International Nuclear Information System (INIS)

Waite, D.A.; Mayberry, J.J.

1986-01-01

The objective of this presentation is to provide an overview of health physics activities anticipated to be required at a high-level nuclear waste repository and to project the numbers of health physics personnel expected to be required to carry out these activities. Health physics personnel receiving consideration in the projections include the health physics manager, shift supervisors, area supervisors, health physicists, and technologists. Phases of the repository addressed are construction, operation, retrieval, and decommissioning. Specific topics discussed in the process of developing the projections are: (a) the basic features of a geologic repository, (b) the staffing requirements of such a repository, (c) health physics involvement in repository operations, and (d) the anticipated schedule for operation of repositories in the United States. A quantitative assessment of future health physics prospects in high-level nuclear waste management is included

An instrumentation and control philosophy for high-level nuclear waste processing facilities

International Nuclear Information System (INIS)

Weigle, D.H.

1990-01-01

The purpose of this paper is to present an instrumentation and control philosophy which may be applied to high-level nuclear waste processing facilities. This philosophy describes the recommended criteria for automatic/manual control, remote/local control, remote/local display, diagnostic instrumentation, interlocks, alarm levels, and redundancy. Due to the hazardous nature of the process constituents of a high-level nuclear waste processing facility, it is imperative that safety and control features required for accident-free operation and maintenance be incorporated. A well-instrumented and controlled process, while initially more expensive in capital and design costs, is generally safer and less expensive to operate. When the long term cost savings of a well designed process is coupled with the high savings enjoyed by accident avoidance, the benefits far outweigh the initial capital and design costs

Immobilized low-level waste disposal options configuration study

International Nuclear Information System (INIS)

Mitchell, D.E.

1995-02-01

This report compiles information that supports the eventual conceptual and definitive design of a disposal facility for immobilized low-level waste. The report includes the results of a joint Westinghouse/Fluor Daniel Inc. evaluation of trade-offs for glass manufacturing and product (waste form) disposal. Though recommendations for the preferred manufacturing and disposal option for low-level waste are outside the scope of this document, relative ranking as applied to facility complexity, safety, remote operation concepts and ease of retrieval are addressed

Liquid level measurement in high level nuclear waste slurries

International Nuclear Information System (INIS)

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

1990-01-01

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

Hanford Immobilized Low-Activity Waste Product Acceptance Test Plan

Energy Technology Data Exchange (ETDEWEB)

Peeler, D.

1999-06-22

'The Hanford Site has been used to produce nuclear materials for the U.S. Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during Pu production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The DOE is proceeding with an approach to privatize the treatment and immobilization of Handord''s LAW and HLW.'

Immobilized enzymes: understanding enzyme - surface interactions at the molecular level.

Science.gov (United States)

Hoarau, Marie; Badieyan, Somayesadat; Marsh, E Neil G

2017-11-22

Enzymes immobilized on solid supports have important and industrial and medical applications. However, their uses are limited by the significant reductions in activity and stability that often accompany the immobilization process. Here we review recent advances in our understanding of the molecular level interactions between proteins and supporting surfaces that contribute to changes in stability and activity. This understanding has been facilitated by the application of various surface-sensitive spectroscopic techniques that allow the structure and orientation of enzymes at the solid/liquid interface to be probed, often with monolayer sensitivity. An appreciation of the molecular interactions between enzyme and surface support has allowed the surface chemistry and method of enzyme attachement to be fine-tuned such that activity and stability can be greatly enhanced. These advances suggest that a much wider variety of enzymes may eventually be amenable to immobilization as green catalysts.

Status of plutonium ceramic immobilization processes and immobilization forms

International Nuclear Information System (INIS)

Ebbinghaus, B.B.; Van Konynenburg, R.A.; Vance, E.R.; Jostsons, A.

1996-01-01

Immobilization in a ceramic followed by permanent emplacement in a repository or borehole is one of the alternatives currently being considered by the Fissile Materials Disposition Program for the ultimate disposal of excess weapons-grade plutonium. To make Pu recovery more difficult, radioactive cesium may also be incorporated into the immobilization form. Valuable data are already available for ceramics form R ampersand D efforts to immobilize high-level and mixed wastes. Ceramics have a high capacity for actinides, cesium, and some neutron absorbers. A unique characteristic of ceramics is the existence of mineral analogues found in nature that have demonstrated actinide immobilization over geologic time periods. The ceramic form currently being considered for plutonium disposition is a synthetic rock (SYNROC) material composed primarily of zirconolite (CaZrTi 2 O 7 ), the desired actinide host phase, with lesser amounts of hollandite (BaAl 2 Ti 6 O 16 ) and rutile (TiO 2 ). Alternative actinide host phases are also being considered. These include pyrochlore (Gd 2 Ti 2 O 7 ), zircon (ZrSiO 4 ), and monazite (CePO 4 ), to name a few of the most promising. R ampersand D activities to address important technical issues are discussed. Primarily these include moderate scale hot press fabrications with plutonium, direct loading of PuO 2 powder, cold press and sinter fabrication methods, and immobilization form formulation issues

Institute for Nuclear Waste Disposal. Annual Report 2011

International Nuclear Information System (INIS)

Geckeis, H.; Stumpf, T.

2012-01-01

The R and D at the Institute for Nuclear Waste Disposal, INE, (Institut fuer Nukleare Entsorgung) of the Karlsruhe Institute of Technology (KIT) focuses on (i) long term safety research for nuclear waste disposal, (ii) immobilization of high level radioactive waste (HLW), (iii) separation of minor actinides from HLW and (iv) radiation protection.

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Involvement of α(2)-adrenergic receptor in the regulation of the blood glucose level induced by immobilization stress.

Science.gov (United States)

Kang, Yu-Jung; Sim, Yun-Beom; Park, Soo-Hyun; Sharma, Naveen; Suh, Hong-Won

2015-01-01

The blood glucose profiles were characterized after mice were forced into immobilization stress with various exposure durations. The blood glucose level was significantly enhanced by immobilization stress for 30 min or 1 h, respectively. On the other hand, the blood glucose level was not affected in the groups which were forced into immobilization stress for 2 or 4 h. We further examined the effect of yohimbine (an α2-adrenergic receptor antagonist) administered systemically or centrally in the immobilization stress model. Mice were pretreated intraperitoneally (i.p.; from 0.5 to 5 mg/kg), intracerebroventricularly (i.c.v.; from 1 to 10 µg/5 µl), or intrathecally (i.t.; from 1 to 10 µg/5 µl) with yohimbine for 10 min and then, forced into immobilization stress for 30 min. The blood glucose level was measured right after immobilization stress. We found that up-regulation of the blood glucose level induced by immobilization stress was abolished by i.p. pretreatment with yohimbine. And the immobilization stress-induced blood glucose level was not inhibited by i.c.v. or i.t. pretreatment with yohimbine at a lower dose (1 µg/5 µl). However, immobilization stress-induced blood glucose level was significantly inhibited by i.c.v. or i.t. pretreatment with yohimbine at higher doses (5 and 10 µg/5 µl). In addition, the i.p. (5 mg/kg), i.c.v. (10 µg/5 µl), or i.t. (10 µg/5 µl) pretreatment with yohimbine reduced hypothalamic glucose transporter 4 expression. The involvement of α2-adrenergic receptor in regulation of immobilization stress- induced blood glucose level was further confirmed by the i.p, i.c.v, or i.t pretreatment with idazoxan, another specific α2-adrenergic receptor antagonist. Finally, i.p., i.c.v., or i.t. pretreatment with yohimbine attenuated the blood glucose level in D-glucose-fed model. We suggest that α2-adrenergic receptors located at the peripheral, the brain and the spinal cord play important roles in the up

Compounds with magnetoplumbite or SLNA type structure as materials for nuclear waste immobilization

International Nuclear Information System (INIS)

Thery, J.; Vivien, D.; Lejus, A.M.; Collongues, R.

1985-01-01

Magnetoplumbite-like structure, and related phases (sodium-lanthanide aluminates: SLNA) are able to accommodate a wide range of elements with various charges and ionic radii. The available coordinences are 4, 5 or 6 for the small cations and 9 or 12 for the large ones. This kind of compounds, which in addition present good chemical inertia, could possibly be used for the immobilization of nuclear waste [fr

Waste incineration and immobilization for nuclear facilities. Status report, October 1977--March 1978

International Nuclear Information System (INIS)

Johnson, A.J.; Burkhardt, S.C.; Ledford, J.A.; Williams, P.M.

1979-01-01

Fluidized bed incineration and processes for immobilization of wastes generated at nuclear facilities are undergoing development. After minor piping modifications to eliminate dust collecting points, a pilot plant fluidized bed incinerator run of 225 continuous hours was successfully completed in a demonstration of component reliability. Vitrification of incinerator ash and other wastes is now being accomplished using a pilot scale unit developed as a continuous flow process

Reduction and resource recycling of high-level radioactive wastes through nuclear transmutation with PHITS code

International Nuclear Information System (INIS)

Fujita, Reiko

2017-01-01

In the ImPACT program of the Cabinet Office, programs are underway to reduce long-lived fission products (LLFP) contained in high-level radioactive waste through nuclear transmutation, or to recycle/utilize useful nuclear species. This paper outlines this program and describes recent achievements. This program consists of five projects: (1) separation/recovery technology, (2) acquisition of nuclear transmutation data, (3) nuclear reaction theory model and simulation, (4) novel nuclear reaction control and development of elemental technology, and (5) discussions on process concept. The project (1) develops a technology for dissolving vitrified solid, a technology for recovering LLFP from high-level waste liquid, and a technology for separating odd and even lasers. Project (2) acquires the new nuclear reaction data of Pd-107, Zr-93, Se-79, and Cs-135 using RIKEN's RIBF or JAEA's J-PARC. Project (3) improves new nuclear reaction theory and structural model using the nuclear reaction data measured in (2), improves/upgrades nuclear reaction simulation code PHITS, and proposes a promising nuclear transmutation pathway. Project (4) develops an accelerator that realizes the proposed transmutation route and its elemental technology. Project (5) performs the conceptual design of the process to realize (1) to (4), and constructs the scenario of reducing/utilizing high-level radioactive waste to realize this design. (A.O.)

Status of plutonium ceramic immobilization processes and immobilization forms

Energy Technology Data Exchange (ETDEWEB)

Ebbinghaus, B.B.; Van Konynenburg, R.A. [Lawrence Livermore National Lab., CA (United States); Vance, E.R.; Jostsons, A. [Australian Nuclear Science and Technology Organization, Menai (Australia)] [and others

1996-05-01

Immobilization in a ceramic followed by permanent emplacement in a repository or borehole is one of the alternatives currently being considered by the Fissile Materials Disposition Program for the ultimate disposal of excess weapons-grade plutonium. To make Pu recovery more difficult, radioactive cesium may also be incorporated into the immobilization form. Valuable data are already available for ceramics form R&D efforts to immobilize high-level and mixed wastes. Ceramics have a high capacity for actinides, cesium, and some neutron absorbers. A unique characteristic of ceramics is the existence of mineral analogues found in nature that have demonstrated actinide immobilization over geologic time periods. The ceramic form currently being considered for plutonium disposition is a synthetic rock (SYNROC) material composed primarily of zirconolite (CaZrTi{sub 2}O{sub 7}), the desired actinide host phase, with lesser amounts of hollandite (BaAl{sub 2}Ti{sub 6}O{sub 16}) and rutile (TiO{sub 2}). Alternative actinide host phases are also being considered. These include pyrochlore (Gd{sub 2}Ti{sub 2}O{sub 7}), zircon (ZrSiO{sub 4}), and monazite (CePO{sub 4}), to name a few of the most promising. R&D activities to address important technical issues are discussed. Primarily these include moderate scale hot press fabrications with plutonium, direct loading of PuO{sub 2} powder, cold press and sinter fabrication methods, and immobilization form formulation issues.

Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia Martinez, H

1999-07-01

In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 {sup L}ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear Triga Mark III

Energy Technology Data Exchange (ETDEWEB)

Garcia M, H.; Emeterio H, M.; Canizal S, C. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, C.P. 11801 Mexico D.F. (Mexico)

2000-07-01

This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

Hanford immobilized low-activity tank waste performance assessment

International Nuclear Information System (INIS)

Mann, F.M.

1998-01-01

The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis

Hanford immobilized low-activity tank waste performance assessment

Energy Technology Data Exchange (ETDEWEB)

Mann, F.M.

1998-03-26

The Hanford Immobilized Low-Activity Tank Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-level fraction of waste presently contained in Hanford Site tanks. The tank waste is the by-product of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste has been stored in underground single and double-shell tanks. The tank waste is to be retrieved, separated into low and high-activity fractions, and then immobilized by private vendors. The US Department of Energy (DOE) will receive the vitrified waste from private vendors and plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at Hanford until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to issue a Disposal Authorization Statement that would allow the modification of the four existing concrete disposal vaults to provide better access for emplacement of the immobilized low-activity waste (ILAW) containers; filling of the modified vaults with the approximately 5,000 ILAW containers and filler material with the intent to dispose of the containers; construction of the first set of next-generation disposal facilities. The performance assessment activity will continue beyond this assessment. The activity will collect additional data on the geotechnical features of the disposal sites, the disposal facility design and construction, and the long-term performance of the waste. Better estimates of long-term performance will be produced and reviewed on a regular basis. Performance assessments supporting closure of filled facilities will be issued seeking approval of those actions necessary to conclude active disposal facility operations. This report also analyzes the long-term performance of the currently planned disposal system as a basis

Immobilization of industrial waste in cement–bentonite clay matrix

Indian Academy of Sciences (India)

Unknown

Immobilization of industrial waste in cement–bentonite clay matrix. I B PLECAS* and S ... high structural integrity and minimizing the risk of escape by leaching. ..... Radioactive Waste Management and Nuclear Fuel Cycle 14. 195. Plecas I ...

A survey of methods for separating and immobilizing krypton-85 arising from a nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Taylor, P.

1990-12-01

This report reviews the literature on methods to separate and immobilize krypton-85 arising from dissolution or prior treatment of nuclear fuel in a reprocessing plant. It was prepared as part of a broader review of fuel reprocessing waste management methods that might find future applications in Canada. Cryogenic distillation is the most fully demonstrated method of separation of krypton from off-gases, but it is complex. In particular, it requires pretreatment of the gas stream to eliminate several other components before the final distillation. The most highly developed alternative process is fluorocarbon adsorption, while several other processes have been investigated on a bench scale. The simplest method of storing radioactive krypton is in compressed-gas cylinders, but the risks of accidental release are increased by the corrosive nature of the decay product, rubidium. Encapsulation in either a metal matrix or a hydrothermally vitrified zeolite appears to offer the most secure immobilization of krypton. Processes for both types of material have been demonstrated inactively on a scale approaching that required for treatment of off-gases from a commercial-scale fuel reprocessing plant. Low-operating temperatures and pressures of the metal encapsulation process, compared with encapsulation in zeolites, represent a definite advantage, but electrical power requirements for the former process are relatively high. It appears that suitable technology is available for separation and immobilization of radioactive krypton, should the need arise in Canada in the future

Nuclear waste management. Quarterly progress report, January-March 1980

Energy Technology Data Exchange (ETDEWEB)

Platt, A.M.; Powell, J.A. (comps.)

1980-06-01

Reported are: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions, engineered barriers, criteria for defining waste isolation, and spent fuel and pool component integrity. (DLC)

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

Science.gov (United States)

Boatner, Lynn A.; Sales, Brian C.

1989-01-01

Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

Hanford Immobilized Low Activity Waste (ILAW) Performance Assessment 2001 Version [Formerly DOE/RL-97-69] [SEC 1 & 2

Energy Technology Data Exchange (ETDEWEB)

MANN, F.M.

2000-08-01

The Hanford Immobilized Low-Activity Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-activity fraction of waste presently contained in Hanford Site tanks. The tank waste is the byproduct of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste is stored in underground single- and double-shell tanks. The tank waste is to be retrieved, separated into low-activity and high-level fractions, and then immobilized by vitrification. The US. Department of Energy (DOE) plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at the Hanford Site until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to modify the current Disposal Authorization Statement for the Hanford Site that would allow the following: construction of disposal trenches; and filling of these trenches with ILAW containers and filler material with the intent to dispose of the containers.

Geology of high-level nuclear waste disposal

International Nuclear Information System (INIS)

Roxburgh, I.S.

1988-01-01

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

Immobilization of chloride-rich radioactive wastes produced by pyrochemical operations

International Nuclear Information System (INIS)

McDaniel, E.W.; Terry, J.W.

1997-08-01

A a result of its former role as a producer of nuclear weapons components, the Rocky Flats Environmental Technology Site (RFETS), Golden, Colorado accumulated a variety of plutonium-contaminated materials. When the level of contamination exceeded a predetermined level (the economic discard limit), the materials were classified as residues rather than waste and were stored for later recovery of the plutonium. Although large quantities of residues were processed, others, primarily those more difficult to process, remain in storage at the site. It is planned for the residues with lower concentrations of plutonium to be disposed of as wastes at an appropriate disposal facility, probably the Waste Isolation Pilot Plant (WIPP). Because the plutonium concentration is too high or because the physical or chemical form would be difficult to get into a form acceptable to WIPP, it may not be possible to dispose of a portion of the residues at WIPP. The pyrochemical salts are among the residues that are difficult to dispose of. For a large percentage of the pyrochemical salts, safeguards controls are required, but WIPP was not designed to accommodate safeguards controls. A potential solution would be to immobilize the salts. These immobilized salts would contain substantially higher plutonium concentrations than is currently permissible but would be suitable for disposal at WIPP. This document presents the results of a review of three immobilization technologies to determine if mature technologies exist that would be suitable to immobilize pyrochemical salts: cement-based stabilization, low-temperature vitrification, and polymer encapsulation. The authors recommend that flow sheets and life-cycle costs be developed for cement-based and low-temperature glass immobilization

Robotics and remote handling concepts for disposal of high-level nuclear waste

International Nuclear Information System (INIS)

McAffee, Douglas; Raczka, Norman; Schwartztrauber, Keith

1997-01-01

This paper summarizes preliminary remote handling and robotic concepts being developed as part of the US Department of Energy's (DOE) Yucca Mountain Project. The DOE is currently evaluating the Yucca Mountain Nevada site for suitability as a possible underground geologic repository for the disposal of high level nuclear waste. The current advanced conceptual design calls for the disposal of more than 12,000 high level nuclear waste packages within a 225 km underground network of tunnels and emplacement drifts. Many of the waste packages may weigh as much as 66 tonnes and measure 1.8 m in diameter and 5.6 m long. The waste packages will emit significant levels of radiation and heat. Therefore, remote handling is a cornerstone of the repository design and operating concepts. This paper discusses potential applications areas for robotics and remote handling technologies within the subsurface repository. It also summarizes the findings of a preliminary technology survey which reviewed available robotic and remote handling technologies developed within the nuclear, mining, rail and industrial robotics and automation industries, and at national laboratories, universities, and related research institutions and government agencies

TWRS retrieval and storage mission. Immobilized low-activity waste disposal plan

International Nuclear Information System (INIS)

Shade, J.W.

1998-01-01

The TWRS mission is to store, treat, and immobilize highly radioactive Hanford waste (current and future tank waste and the encapsulated cesium and strontium) in a safe, environmentally sound, and cost-effective manner (TWRS JMN Justification for mission need). The mission includes retrieval, pretreatment, immobilization, interim storage and disposal, and tank closure. As part of this mission, DOE has established the TWRS Office to manage all Hanford Site tank waste activities. The TWRS program has identified the need to store, treat, immobilize, and dispose of the highly radioactive Hanford Site tank waste and encapsulated cesium and strontium materials in an environmentally sound, safe, and cost-effective manner. To support environmental remediation and restoration at the Hanford Site a two-phase approach to using private contractors to treat and immobilize the low-activity and high-level waste currently stored in underground tanks is planned. The request for proposals (RFP) for the first phase of waste treatment and immobilization was issued in February 1996 (Wagoner 1996) and initial contracts for two private contractor teams led by British Nuclear Fuels Ltd. and Lockheed-Martin Advanced Environmental Services were signed in September 1996. Phase 1 is a proof-of-concept and commercial demonstration effort to demonstrate the technical and business feasibility of using private facilities to treat Hanford Site waste, maintain radiological, nuclear, process, and occupational safety; and maintain environmental protection and compliance while reducing lifecycle costs and waste treatment times. Phase 1 production of ILAW is planned to begin in June 2002 and could treat up to about 13 percent of the waste. Phase 1 production is expected to be completed in 2007 for minimum order quantities or 2011 for maximum order quantities. Phase 2 is a full-scale production effort that will begin after Phase 1 and treat and immobilize most of the waste. Phase 2 production is

Application of SYNROC to high-level defense wastes

International Nuclear Information System (INIS)

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

1981-01-01

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

German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

Directory of Open Access Journals (Sweden)

A. Schwenk-Ferrero

2013-01-01

Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

The scope and nature of the problem of high level nuclear waste disposal

International Nuclear Information System (INIS)

Jennekens, J.

1981-09-01

The disposal of high level nuclear waste poses a challenge to the Canadian technical and scientific communities, but a much greater challenge to government and industry leaders who must convince the public that the so-called 'problem' can be resolved by a pragmatic approach utilizing existing skills and knowledge. This paper outlines the objectives of radioactive waste management, the quantities of high level waste expected to be produced by the Canadian nuclear power program, the regulatory process which will apply and the government initiatives which have been and will be taken to ensure that the health, safety, security, and environmental interests of the public will be protected. (author)

Plutonium immobilization in glass and ceramics

Energy Technology Data Exchange (ETDEWEB)

Knecht, D.A. [Lockheed Martin Idaho Technologies, Idaho Falls (United States); Murphy, W.M. [Southwest Research Institute, San Antonio, TX (United States)

1996-05-01

The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposium papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 {degrees}C, a higher temperature (1450 {degrees}C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature.

Plutonium immobilization in glass and ceramics

International Nuclear Information System (INIS)

Knecht, D.A.; Murphy, W.M.

1996-01-01

The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposium papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 degrees C, a higher temperature (1450 degrees C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature

Immobilization of Uranium Silicides in Sintered Glass

International Nuclear Information System (INIS)

Mateos, P.; Russo, D.O.; Heredia, A.D.; Sanfilippo, M.

2003-01-01

High activity nuclear spent fuels vitrification by fusion is a well known technology which has industrial scale in France, England, Japan, EEUU. Borosilicates glasses are used in this process.Sintered glasses are an alternative to the immobilization task in which there is also a wide experience around the world.The available technics are: cold pressing and sintering , hot-pressing and hot isostatic pressing.This work compares Borosilicates and Iron silicates sintered glasses behaviour when different ammounts of nuclear simulated waste is added

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Fracture toughness measurements on a glass bonded sodalite high-level waste form

International Nuclear Information System (INIS)

DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T. P.

1999-01-01

The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies

Transport of Spent Nuclear Fuels, High and Intermediate Level Wastes: A Continuous Challenge

International Nuclear Information System (INIS)

Otton, C.; Blachet, L.

2009-01-01

For more than 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the used nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. In this presentation we will focus on the casks for the spent fuel, high level waste and intermediate level waste transportation. Answering to the constant evolution of the nuclear industry transport needs is a challenge that TN International faces routinely. Concerning the spent nuclear fuel transportation, TN International has developed in the early 80's a fleet of TN12 type casks fitted with several types of baskets able to safely transport all the spent fuel from the nuclear power plant or the research laboratories to AREVA La Hague plant. The current challenge is the design of a new transport cask generation taking into account the needs of the industry for the next 30 years. The replacement of the TN12 cask generation is to be scheduled as the regulations have changed and the fuel characteristics have evolved. The new generation of casks will take into account all the technical evolutions made during the TN12 thirty years of use. MOX spent fuel has now its dedicated cask: the TN112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in 2008 in the EDF nuclear power plant of Saint-Laurent-des-Eaux. Concerning the high level waste such as the La Hague vitrified residues a whole fleet of casks has been developed such as the TN 28 VT dedicated to transport, the TN81 and TN85 dedicated to transport and storage. These casks have permitted the

Nuclear waste management. Quarterly progress report, October-December 1979

Energy Technology Data Exchange (ETDEWEB)

Platt, A.M.; Powell, J.A. (comps.)

1980-04-01

Progress and activities are reported on the following: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization programs, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, monitoring of unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions technology, spent fuel and fuel pool integrity program, and engineered barriers. (DLC)

Siting high-level nuclear waste repositories: A progress report for Rhode Island

International Nuclear Information System (INIS)

Frohlich, R.K.; Vild, B.F.

1986-03-01

In this booklet, we will not try to argue the pros and cons of nuclear power or weapons production. We will focus instead on the issue of nuclear waste disposal. With the passage of the Nuclear Waste Policy Act (NWPA) of 1982, the US Congress and the President charged federal and state regulators with the responsibility of settling that issue by the end of this century - with extensive public involvement. This booklet, now in its second printing, is designed to explain the nature of ''high-level'' nuclear waste, the essential criteria for its safe and permanent disposal, and Rhode Island's participation in the federal repository program. It has been funded from a USDOE grant derived from a utility-financed Nuclear Waste Fund established under the NWPA. 17 refs., 10 figs., 2 tabs

Properties of Immobilized Candida antarctica Lipase B on Highly Macroporous Copolymer

International Nuclear Information System (INIS)

Handayani, N.; Achmad, S.; Wahyuningrum, D.

2011-01-01

In spite of their excellent catalytic properties, enzymes should be improved before their implementation both in industrial and laboratorium scales. Immobilization of enzyme is one of the ways to improve their properties. Candida antarctica lipase B (Cal-B) has been reported in numerous publications to be a particularly useful enzyme catalizing in many type of reaction including regio- and enantio- synthesis. For this case, cross-linking of immobilized Cal-B with 1,2,7,8 diepoxy octane is one of methods that proved significantly more stable from denaturation by heat, organic solvents, and proteolysis than lyophilized powder or soluble enzymes. More over, the aim of this procedure is to improve the activity and reusability of lipase. Enzyme kinetics test was carried out by transesterification reaction between 4-nitrophenyl acetate (pNPA) and methanol by varying substrate concentrations, and the result is immobilized enzymes follows the Michaelis-Menten models and their activity is match with previous experiment. Based on the V max values, the immobilized enzymes showed higher activity than the free enzyme. Cross-linking of immobilized lipase indicate that cross-linking by lower concentration of cross-linker, FIC (immobilized lipase that was incubated for 24 h) gave the highest activity and cross-linking by higher concentration of cross-linker, PIC (immobilized lipase that was incubated for 2 h) gives the highest activity. However, pore size and saturation level influenced their activity. (author)

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.

Level density from realistic nuclear potentials

International Nuclear Information System (INIS)

Calboreanu, A.

2006-01-01

Nuclear level density of some nuclei is calculated using a realistic set of single particle states (sps). These states are derived from the parameterization of nuclear potentials that describe the observed sps over a large number of nuclei. This approach has the advantage that one can infer level density for nuclei that are inaccessible for a direct study, but are very important in astrophysical processes such as those close to the drip lines. Level densities at high excitation energies are very sensitive to the actual set of sps. The fact that the sps spectrum is finite has extraordinary consequences upon nuclear reaction yields due to the leveling-off of the level density at extremely high excitation energies wrongly attributed so far to other nuclear effects. Single-particle level density parameter a parameter is extracted by fitting the calculated densities to the standard Bethe formula

Using electrochemical separation to reduce the volume of high-level nuclear waste

International Nuclear Information System (INIS)

Slater, S.A.; Gay, E.C.

1998-01-01

Argonne National Laboratory (ANL) has developed an electrochemical separation technique called electrorefining that will treat a variety of metallic spent nuclear fuel and reduce the volume of high-level nuclear waste that requires disposal. As part of that effort, ANL has developed a high throughput electrorefiner (HTER) that has a transport rate approximately three times faster than electrorefiners previously developed at ANL. This higher rate is due to the higher electrode surface area, a shorter transport path, and more efficient mixing, which leads to smaller boundary layers about the electrodes. This higher throughput makes electrorefining an attractive option in treating Department of Energy spent nuclear fuels. Experiments have been done to characterize the HTER, and a simulant metallic fuel has been successfully treated. The HTER design and experimental results is discussed

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Parity dependence of the nuclear level density at high excitation

International Nuclear Information System (INIS)

Rao, B.V.; Agrawal, H.M.

1995-01-01

The basic underlying assumption ρ(l+1, J)=ρ(l, J) in the level density function ρ(U, J, π) has been checked on the basis of high quality data available on individual resonance parameters (E 0 , Γ n , J π ) for s- and p-wave neutrons in contrast to the earlier analysis where information about p-wave resonance parameters was meagre. The missing level estimator based on the partial integration over a Porter-Thomas distribution of neutron reduced widths and the Dyson-Mehta Δ 3 statistic for the level spacing have been used to ascertain that the s- and p-wave resonance level spacings D(0) and D(1) are not in error because of spurious and missing levels. The present work does not validate the tacit assumption ρ(l+1, J)=ρ(l, J) and confirms that the level density depends upon parity at high excitation. The possible implications of the parity dependence of the level density on the results of statistical model calculations of nuclear reaction cross sections as well as on pre-compound emission have been emphasized. (orig.)

Nuclear Waste Vitrification in the U.S.: Recent Developments and Future Options

International Nuclear Information System (INIS)

Vienna, John D.

2010-01-01

Nuclear power plays a key role in maintaining current world wide energy growth while minimizing the greenhouse gas emissions. A disposition path for used nuclear fuel (UNF) must be found for this technology to achieve its promise. One likely option is the recycling of UNF and immobilization of the high-level waste (HLW) by vitrification. Vitrification is the technology of choice for immobilizing HLW from defense and commercial fuel reprocessing around the world. Recent advances in both recycling technology and vitrification show great promise in closing the nuclear fuel cycle in an efficient and economical fashion. This article summarizes the recent trends developments and future options in waste vitrification for both defense waste cleanup and closing the nuclear fuel cycle in the U.S.

High Efficiency Acetylcholinesterase Immobilization on DNA Aptamer Modified Surfaces

Directory of Open Access Journals (Sweden)

Orada Chumphukam

2014-04-01

Full Text Available We report here the in vitro selection of DNA aptamers for electric eel acetylcholinesterase (AChE. One selected aptamer sequence (R15/19 has a high affinity towards the enzyme (Kd = 157 ± 42 pM. Characterization of the aptamer showed its binding is not affected by low ionic strength (~20 mM, however significant reduction in affinity occurred at high ionic strength (~1.2 M. In addition, this aptamer does not inhibit the catalytic activity of AChE that we exploit through immobilization of the DNA on a streptavidin-coated surface. Subsequent immobilization of AChE by the aptamer results in a 4-fold higher catalytic activity when compared to adsorption directly on to plastic.

Radioactive waste immobilization in protective ceramic forms by the HIP method at high pressures

International Nuclear Information System (INIS)

Sayenko, S.Yu.; Kantsedal, V.P.; Tarasov, R.V.; Starchenko, V.A.; Lyubtsev, R.I.

1993-01-01

Intense research activities have been carried out in recent years at the Kharkov Institute of Physics and Technology (KIPT) to develop the method of hot isostatic pressing (HIP) for immobilizing radioactive (primarily, high-level) wastes. With this method, the radioactive material is immobilized in a matrix under the simultaneous action of high pressures (up to 6,000 atm) and appropriate temperatures. The process has 2 variants: (1) radioactive wastes are treated as powders of oxides resulting from calcination during chemical treatment of spent fuel. In this case the radioactive material enters into the crystalline structure of the immobilized matrix or is distributed in the matrix as a homogeneous mixture; (2) protective barrier layers are pressed on spent fuel rods or their pieces as radioactive wastes, by the HIP method (fuel rod encapsulation in a protective form). Based on numerous results from various studies, the authors suggest that various ceramic compositions should be used as protective materials. Here the authors report two trends of their investigations: (1) development of ecologically clean process equipments for radioactive waste treatment by the HIP method; (2) manufacture of promising protective ceramic compositions and investigation of their physico-mechanical properties

Development of anodic stripping voltametry for the determination of palladium in high level nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Bhardwaj, T. K. [North Carolina State University, Raleigh (United States); Sharma, H. S.; Affarwal, S. K. [Bhabha Atomic Research Centre, Mumbai (India); Jain, P. C. [Meerut College, Meerut (India)

2012-12-15

Deposition potential, deposition time, square wave frequency, rotation speed of the rotating disc electrode, and palladium concentration were studied on a Glassy Carbon Electrode (GCE) in 0.01M HCl for the determination of palladium in High Level Nuclear Waste (HLNW) by anodic stripping voltammetry. Experimental conditions were optimized for the determination of palladium at two different, 10-8 and 10-7 M, levels. Error and standard deviation of this method were under 1% for all palladium standard solutions. The developed technique was successfully applied as a subsidiary method for the determination of palladium in simulated high level nuclear waste with very good precision and high accuracy (under 1 % error and standard deviation).

A Strategy for Maintenance of the Long-Term Performance Assessment of Immobilized Low-Activity Waste Glass

Energy Technology Data Exchange (ETDEWEB)

Ryan, Joseph V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Freedman, Vicky L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-09-28

Approximately 50 million gallons of high-level radioactive mixed waste has accumulated in 177 buried single- and double-shell tanks at the Hanford Site in southeastern Washington State as a result of the past production of nuclear materials, primarily for defense uses. The United States Department of Energy (DOE) is proceeding with plans to permanently dispose of this waste. Plans call for separating the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, which will be vitrified at the Hanford Waste Treatment and Immobilization Plant (WTP). Principal radionuclides of concern in LAW are 99Tc, 129I, and U, while non-radioactive contaminants of concern are Cr and nitrate/nitrite. HLW glass will be sent off-site to an undetermined federal site for deep geological disposal while the much larger volume of immobilized low-activity waste will be placed in the on-site, near-surface Integrated Disposal Facility (IDF).

Application of self-propagation high-temperature synthesis for immobilization of hard radioactive wastes in ceramet materials

International Nuclear Information System (INIS)

Ilyin, E.; Pashkeev, I.; Senin, A.; Gerasimova, N.

2001-01-01

The possibility of self-propagating high-temperature synthesis (SPHTS) application for an immobilization of solid high level wastes (HLW) in cermet materials is considered. The schemes of multilayer cermet blocks formation are offered. Such blocks consist of a ceramet core with immobilized HLW and a protective cover - ceramet without HLW. The influence of the base components form (pure Ti and Si, ferrotitanium and ferrosilicon), metallic components (Ni, Cu, Cr, Fe, ferrochromium) and nonmetallic components (SiO 2 , Al 2 O 3 , TiO 2 ) on burning rate and cover ceramet structure is investigated in compositions on a basis of Ti+B, Ti+Si, Ti+C systems. Model samples of multilayer cermet blocks are manufactured using of HLW simulators. (authors)

Sodium zirconium phosphate (NZP) as a host structure for nuclear waste immobilization: A review

International Nuclear Information System (INIS)

Scheetz, B.E.; Agrawal, D.K.; Breval, E.; Roy, R.

1994-01-01

Sodium zirconium phosphate [NZP] structural family, of which NaZr 2 P 3 O 12 is the parent composition, has been reviewed as a host ceramic waste form for nuclear waste immobilization. NZP compounds are characterized for their ionic conductivity, low thermal expansion and structural flexibility to accommodate a large number of multivalent ions. This latter property of the [NZP] structure allows the incorporation of almost all 42 nuclides present in a typical commercial nuclear waste. The leach studies of simulated waste forms based on NZP have shown reasonable resistance for the release of its constituents. The calculation of dissolution rates of NZP structure has demonstrated that it would take 20,000 times longer to dissolved NZP than quartz

Fair rules for siting a high-level nuclear waste repository

International Nuclear Information System (INIS)

Easterling, D.

1992-01-01

Geologic repositories are designed to resolve the ever-growing problem of high-level nuclear waste, but these facilities invite intense local opposition due to the perceived severity of the risks and the possibility of stigma effects. This analysis examines whether the perceived fairness of the siting process affects local residents' support for hosting a repository. In particular, a survey of 1,001 Nevada residents is used to test the hypothesis that an individual's willingness to accept a local repository will increase if he or she is convinced that this is the safest disposal option available. A logistic analysis indicates that beliefs regarding relative suitability have an independent effect on the acceptability of a local repository (i.e., Yucca Mountain). The article then considers the question of how to implement an optimizing strategy for siting facilities, comparing an idealized strategy against the original Nuclear Waste Policy Act (NWPA) of 1982 and the Amendments Act of 1987. Although choosing the safest site seems as if it could enhance public acceptance of the repository program, there is currently little prospect of identifying the best option to the high-level waste problem and, as a results, little chance of gaining the public support that is necessary to promote a successful siting outcome. 81 refs., 1 fig., 5 tabs

Dynamic nuclear polarization at high Landau levels in a quantum point contact

Science.gov (United States)

Fauzi, M. H.; Noorhidayati, A.; Sahdan, M. F.; Sato, K.; Nagase, K.; Hirayama, Y.

2018-05-01

We demonstrate a way to polarize and detect nuclear spin in a gate-defined quantum point contact operating at high Landau levels. Resistively detected nuclear magnetic resonance (RDNMR) can be achieved up to the fifth Landau level and at a magnetic field lower than 1 T. We are able to retain the RDNMR signals in a condition where the spin degeneracy of the first one-dimensional (1D) subband is still preserved. Furthermore, the effects of orbital motion on the first 1D subband can be made smaller than those due to electrostatic confinement. This developed RDNMR technique is a promising means to study electronic states in a quantum point contact near zero magnetic field.

Immobilization of iodine in concrete

Science.gov (United States)

Clark, Walter E.; Thompson, Clarence T.

1977-04-12

A method for immobilizing fission product radioactive iodine recovered from irradiated nuclear fuel comprises combining material comprising water, Portland cement and about 3-20 wt. % iodine as Ba(IO.sub.3).sub.2 to provide a fluid mixture and allowing the fluid mixture to harden, said Ba(IO.sub.3).sub.2 comprising said radioactive iodine. An article for solid waste disposal comprises concrete prepared by this method. BACKGROUND OF THE INVENTION This invention was made in the course of, or under a contract with the Energy Research and Development Administration. It relates in general to reactor waste solidification and more specifically to the immobilization of fission product radioactive iodine recovered from irradiated nuclear fuel for underground storage.

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

International Nuclear Information System (INIS)

Wicks, G.G.

1980-01-01

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

Comparison of bitumen and cement immobilization of intermediate- and low-level radioactive waste

International Nuclear Information System (INIS)

Voss, J.W.

1979-01-01

This paper discusses a systems comparison of two available immobilization processes for intermediate- and low-level radioactive wastes -- bitumen and cement. This study examines a conceptual coprocessed UO 2 - PuO 2 fuel cycle. Radioactive wastes are generated at each stage of this fuel cycle. This study focuses on these transuranic (TRU) wastes generated at a conceptual Fuel Coprocessing Facility. In this report, these wastes are quantified, the immobilization systems conceptualized to process these wastes are presented, and a comparison of the systems is made

Immobilization of wet solid wastes at nuclear power plants

International Nuclear Information System (INIS)

Neilson, R.M. Jr.

1977-01-01

Wet solid wastes are classified into four basic types: spent resins, filter sludges, evaporator concentrates, and miscellaneous liquids. Although the immobilization of wet solid wastes is primarily concerned with the incorporation of the waste with a solidification agent, there are a number of other discrete operations or subsystems involved in the treatment of these wastes that may affect the immobilized waste product. The immobilization process may be broken down into five basic operations: waste collection, waste pretreatment, solidification agent handling, mixing/packaging, and waste package handling. The properties of the waste forms that are ultimately shipped from the reactor site are primarily influenced by the methods utilized during the waste collection, waste pretreatment and mixing/packaging operations. The mixing/packaging (solidification) operation is perhaps the most important stage of the immobilization process. The basic solidification agent types are: absorbants, hydraulic cement, urea-formaldehyde, bitumen, and other polymer systems

Should the U.S. proceed to consider licensing deep geological disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Curtiss, J.R.

1993-01-01

The United States, as well as other countries facing the question of how to handle high-level nuclear waste, has decided that the most appropriate means of disposal is in a deep geologic repository. In recent years, the Radioactive Waste Management Committee of the Nuclear Energy Agency has developed several position papers on the technical achievability of deep geologic disposal, thus demonstrating the serious consideration of deep geologic disposal in the international community. The Committee has not, as yet, formally endorsed disposal in a deep geologic repository as the preferred method of handling high-level nuclear waste. The United States, on the other hand, has studied the various methods of disposing of high-level nuclear waste, and has determined that deep geologic disposal is the method that should be developed. The purpose of this paper is to present a review of the United States' decision on selecting deep geologic disposal as the preferred method of addressing the high-level waste problem. It presents a short history of the steps taken by the U.S. in determining what method to use, discusses the NRC's waste Confidence Decision, and provides information on other issues in the U.S. program such as reconsideration of the final disposal standard and the growing inventory of spent fuel in storage

Immobilization of Radioactive Rare Earth oxide Waste by Solid Phase Sintering

International Nuclear Information System (INIS)

Ahn, Byung Gil; Park, Hwan Seo; Kim, Hwan Young; Lee, Han Soo; Kim, In Tae

2010-01-01

In the pyroprocessing of spent nuclear fuels, LiCl-KCl waste salt containing radioactive rare earth chlorides are generated. The radioactive rare earth oxides are recovered by co-oxidative precipitation of rare earth elements. The powder phase of rare earth oxide waste must be immobilized to produce a monolithic wasteform suitable for storage and ultimate disposal. The immobilization of these waste developed in this study involves a solid state sintering of the waste with host borosilicate glass and zinc titanate based ceramic matrix (ZIT). And the rare-earth monazite which synthesised by reaction of ammonium di-hydrogen phosphate with the rare earth oxides waste, were immobilized with the borosilicate glass. It is shown that the developed ZIT ceramic wasteform is highly resistant the leaching process, high density and thermal conductivity.

Chem I Supplement. Chemistry Related to Isolation of High-Level Nuclear Waste.

Science.gov (United States)

Hoffman, Darleane C.; Choppin, Gregory R.

1986-01-01

Discusses some of the problems associated with the safe disposal of high-level nuclear wastes. Describes several waste disposal plans developed by various nations. Outlines the multiple-barrier concept of isolation in deep geological questions associated with the implementation of such a method. (TW)

Development, evaluation, and selection of candidate high-level waste forms

International Nuclear Information System (INIS)

Bernadzikowski, T.A.; Allender, J.S.; Gordon, D.E.; Gould, T.H. Jr.

1982-01-01

The seven candidate waste forms, evaluated as potential media for the immobilization and gelogic disposal of high-level nuclear wastes were borosilicate glass, SYNROC, tailored ceramic, high-silica glass, FUETAP concrete, coated sol-gel particles, and glass marbles in a lead matrix. The evaluation, completed on August 1, 1981, combined preliminary waste form evaluations conducted at Department of Energy (DOE) defense waste-sites and at independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate-based ceramic, SYNROC, were selected as the reference and alternative forms, respectively, for continued development and evaluation in the National HLW Program. The borosilicate glass and ceramic forms were further compared during FY-1982 on the basis of risk assessments, cost comparisons, properties comparisons, and conformance with proposed regulatory and repository criteria. Both the glass and ceramic forms are viable candidates for use at DOE defense HLW sites; they are also candidates for immobilization of commercial reprocessing wastes. This paper describes the waste form screening process, discusses each of the four major inputs considered in the selection of the two forms in 1981, and presents a brief summary of the comparisons of the two forms during 1982 and the selection process to determine the final form for SRP defense HLW

UK-Russian collaboration high level waste immobilization studies

International Nuclear Information System (INIS)

Anderson, E.; Burakov, B.; Galkin, B.; Starchenko, V.; Vasiliev, V.; Shulyak, N.; Homes, R.G.G.; Weaver, W.; Goddard, D.; Clegg, R.; Richardson, S.

1995-01-01

Recent social changes in Russia have opened up many opportunities for business collaboration. To build on this, in 1992 British Nuclear Fuels Limited (BNFL) concluded and signed an agreement with the Russian ministry MINATOM to collaborate on a wide range of topics relevant to the international nuclear industry. These covered the such subjects as developing national regulatory frameworks, sharing operational experiences and practices, and establishing collaborative R and D programmes. One outcome of the agreement with MINATOM has been the setting up of a number of collaborative R and D projects with the V.G. Kholpin Radium Institute in St. Petersburg. This paper presents the results from one of these joint programmes, and describes the mutual benefits that can be obtained from such collaborative work. (authors)

Design and operating features of the high-level waste vitrification system for the West Valley demonstration project

International Nuclear Information System (INIS)

Siemens, D.H.; Beary, M.M.; Barnes, S.M.; Berger, D.N.; Brouns, R.A.; Chapman, C.C.; Jones, R.M.; Peters, R.D.; Peterson, M.E.

1986-03-01

A liquid-fed joule-heated ceramic melter system is the reference process for immobilization of the high-level liquid waste in the US and several foreign countries. This system has been under development for over ten years at Pacific Northwest Laboratory and other national laboratories operated for the US Department of Energy. Pacific Northwest Laboratory contributed to this research through its Nuclear Waste Treatment Program and used applicable data to design and test melters and related systems using remote handling of simulated radioactive wastes. This report describes the equipment designed in support of the high-level waste vitrification program at West Valley, New York. Pacific Northwest Laboratory worked closely with West Valley Nuclear Services Company to design a liquid-fed ceramic melter, a liquid waste preparation and feed tank and pump, an off-gas treatment scrubber, and an enclosed turntable for positioning the waste canisters. Details of these designs are presented including the rationale for the design features and the alternatives considered

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

International Nuclear Information System (INIS)

Vance, E.R.

1980-01-01

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

Plasma Treated High-Density Polyethylene (HDPE Medpor Implant Immobilized with rhBMP-2 for Improving the Bone Regeneration

Directory of Open Access Journals (Sweden)

Jin-Su Lim

2014-01-01

Full Text Available We investigate the bone generation capacity of recombinant human bone morphogenetic protein-2 (rhBMP-2 immobilized Medpor surface through acrylic acid plasma-polymerization. Plasma-polymerization was carried out at a 20 W at an acrylic acid flow rate of 7 sccm for 5 min. The plasma-polymerized Medpor surface showed hydrophilic properties and possessed a high density of carboxyl groups. The rhBMP-2 was immobilized with covalently attached carboxyl groups using 1-ethyl-3-(3-dimethylaminopropyl carbodiimide and N-hydroxysuccinimide. Carboxyl groups and rhBMP-2 immobilization on the Medpor surface were identified by Fourier transform infrared spectroscopy. The activity of Medpor with rhBMP-2 immobilized was examined using an alkaline phosphatase assay on MC3T3-E1 cultured Medpor. These results showed that the rhBMP-2 immobilized Medpor increased the level of MC3T3-E1 cell differentiation. These results demonstrated that plasma surface modification has the potential to immobilize rhBMP-2 on polymer implant such as Medpor and can be used for the binding of bioactive nanomolecules in bone tissue engineering.

Plutonium Disposition by Immobilization

International Nuclear Information System (INIS)

Gould, T.; DiSabatino, A.; Mitchell, M.

2000-01-01

The ultimate goal of the Department of Energy (DOE) Immobilization Project is to develop, construct, and operate facilities that will immobilize between 17 to 50 tonnes (MT) of U.S. surplus weapons-usable plutonium materials in waste forms that meet the ''spent fuel'' standard and are acceptable for disposal in a geologic repository. Using the ceramic can-in-canister technology selected for immobilization, surplus plutonium materials will be chemically combined into ceramic forms which will be encapsulated within large canisters of high level waste (HLW) glass. Deployment of the immobilization capability should occur by 2008 and be completed within 10 years. In support of this goal, the DOE Office of Fissile Materials Disposition (MD) is conducting development and testing (D and T) activities at four DOE laboratories under the technical leadership of Lawrence Livermore National Laboratory (LLNL). The Savannah River Site has been selected as the site for the planned Plutonium Immobilization Plant (PIP). The D and T effort, now in its third year, will establish the technical bases for the design, construction, and operation of the U. S. capability to immobilize surplus plutonium in a suitable and cost-effective manner. Based on the D and T effort and on the development of a conceptual design of the PIP, automation is expected to play a key role in the design and operation of the Immobilization Plant. Automation and remote handling are needed to achieve required dose reduction and to enhance operational efficiency

Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

Energy Technology Data Exchange (ETDEWEB)

SK Sundaram; ML Elliott; D Bickford

1999-11-19

SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

International Nuclear Information System (INIS)

Sundaram, S.K.; Elliott, M.L.; Bickford, D.

1999-01-01

SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described

Waste incineration and immobilization for nuclear facilities. Status report, April-September 1978

International Nuclear Information System (INIS)

Johnson, A.J.; Williams, P.M.; Burkhardt, S.C.; Ledford, J.A.; Gallagher, K.Y.

1980-01-01

The fluidized bed incinerator and waste immobilization processes are being developed to process various liquid and solid wastes that are generated by a nuclear facility. The versatility of the incinerator liquid waste handling system has been enhanced by recent changes made in the pumping and related piping system. Tributyl phosphate-solvent incineration has been evaluated thoroughly using the pilot plant fluidized bed incinerator. Vitrified glass pellets were made to determine operating parameters of a resistance-heated reactor and to produce samples for testing. Procedures were developed for testing the product pellets. A simplified start-up procedure was devised as development continued on a second type of reactor, the Joule-heated melter

Status of the United States' high-level nuclear waste disposal program

International Nuclear Information System (INIS)

Rusche, B.

1985-01-01

The Nuclear Waste Policy Act of 1982 is a remarkable piece of legislation in that there is general agreement on its key provisions. Nevertheless, this is a program intended to span more than a century, with some choices by Congress, states, Indian tribes and the nuclear power industry yet to be made. The crafters of the Act clearly recognized this. And further, the crafters recognized ''. . .that. . .state, Indian tribe and public participation in the planning and development of repositories is essential in order to promote public confidence in the safety of disposal of such waste and spent fuel . . . High-level radioactive waste and spent nuclear fuel have become major subjects of public concern, and appropriate precautions must be taken to ensure that such waste and spent fuel do not adversely affect the public health and safety and the environment for this or future generations

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-08-15

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

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

International Nuclear Information System (INIS)

Dees, L.A.

1994-01-01

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

Facile and high-efficient immobilization of histidine-tagged multimeric protein G on magnetic nanoparticles

Science.gov (United States)

Lee, Jiho; Chang, Jeong Ho

2014-12-01

This work reports the high-efficient and one-step immobilization of multimeric protein G on magnetic nanoparticles. The histidine-tagged (His-tag) recombinant multimeric protein G was overexpressed in Escherichia coli BL21 by the repeated linking of protein G monomers with a flexible linker. High-efficient immobilization on magnetic nanoparticles was demonstrated by two different preparation methods through the amino-silane and chloro-silane functionalization on silica-coated magnetic nanoparticles. Three kinds of multimeric protein G such as His-tag monomer, dimer, and trimer were tested for immobilization efficiency. For these tests, bicinchoninic acid (BCA) assay was employed to determine the amount of immobilized His-tag multimeric protein G. The result showed that the immobilization efficiency of the His-tag multimeric protein G of the monomer, dimer, and trimer was increased with the use of chloro-silane-functionalized magnetic nanoparticles in the range of 98% to 99%, rather than the use of amino-silane-functionalized magnetic nanoparticles in the range of 55% to 77%, respectively.

Review of high-level waste form properties

International Nuclear Information System (INIS)

Rusin, J.M.

1980-12-01

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

Numerical investigation of high level nuclear waste disposal in deep anisotropic geologic repositories

KAUST Repository

Salama, Amgad; El Amin, Mohamed F.; Sun, Shuyu

2015-01-01

One of the techniques that have been proposed to dispose high level nuclear waste (HLW) has been to bury them in deep geologic formations, which offer relatively enough space to accommodate the large volume of HLW accumulated over the years since

Studies on gelation of sodium silicate hydrosol for immobilization of high level liquid waste (HLLW).

Energy Technology Data Exchange (ETDEWEB)

Abdel Raouf, M W [Hot Lab. Centre, Atomic Energy Authority, Cairo (Egypt); Sharaf El-deen, A N; El-Dessouky, M M [Military Technical College, Kobry El-Kobbah, Cairo (Egypt)

1995-10-01

Immobilization of the simulated high-level liquid waste (HLLW) was performed via the gelation with sodium silicate hydrosol at room temperature. The simulated waste in this study, was represented by the electrolytes of Li, Na, K, Cs, Co and Sr at different concentrations. Specific loading of the liquid waste with 0.6 M Mg (NO{sub 3})2 and tailoring with Al salts were tried during most of the gelation processes. Mineral acid (HCl or {sub 3}) were added during the gelation processes to achieve the gel point, especially when lower concentrations of the simulated waste were used. The obtained hydrogel were dried to obtain the solid gel form. The gelation processes were investigated in terms of the different factors that affected them, namely: temperature, pH, changes in the concentration of the initial hydrosol and the used electrolytes. The efficiency of the gelation processes was investigated from the ratio of the amount of simulated waste reacted (m mole) to the initial silicate used (m mole), i.e. X value. Lower X values were observed when using multi valent cations (higher polarizing power). A special effect of increasing the sorption of metal cations in the silica matrix was observed when Al{sup 3+} replaced Si{sup 4+} in the three-dimensional network structure of the matrix. 3 figs., 7 tabs.

A method for assay of special nuclear material in high level liquid waste streams

International Nuclear Information System (INIS)

Venkata Subramani, C.R.; Swaminathan, K.; Asuvathraman, R.; Kutty, K.V.G.

2003-01-01

The assay of special nuclear material in the high level liquid waste streams assumes importance as this is the first stage in the extraction cycle and considerable losses of plutonium could occur here. This stream contains all the fission products as also the minor actinides and hence normal nuclear techniques cannot be used without prior separation of the special nuclear material. This paper presents the preliminary results carried out using wavelength dispersive x-ray fluorescence as part of the developmental efforts to assay SNM in these streams by instrumental techniques. (author)

Technetium Incorporation in Glass for the Hanford Tank Waste Treatment and Immobilization Plant

Energy Technology Data Exchange (ETDEWEB)

Kruger, Albert A.; Kim, Dong Sang

2015-01-14

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of 99Tc (~9 × 10E2 TBq or ~2.5 × 104 Ci or ~1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with 99Tc is its high mobility in addition to a long half-life (2.1×105 yrs). The pertechnetate ion (TcO4-) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater

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

International Nuclear Information System (INIS)

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

1997-01-01

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

SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

Energy Technology Data Exchange (ETDEWEB)

Amoroso, J. [Savannah River National Laboratory, Aiken, SC (United States); Marra, J. [Savannah River National Laboratory, Aiken, SC (United States)

2014-10-02

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

International Nuclear Information System (INIS)

Amoroso, J.; Marra, J.

2014-01-01

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear fuel. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing

Physical modeling of joule heated ceramic glass melters for high level waste immobilization

International Nuclear Information System (INIS)

Quigley, M.S.; Kreid, D.K.

1979-03-01

This study developed physical modeling techniques and apparatus suitable for experimental analysis of joule heated ceramic glass melters designed for immobilizing high level waste. The physical modeling experiments can give qualitative insight into the design and operation of prototype furnaces and, if properly verified with prototype data, the physical models could be used for quantitative analysis of specific furnaces. Based on evaluation of the results of this study, it is recommended that the following actions and investigations be undertaken: It was not shown that the isothermal boundary conditions imposed by this study established prototypic heat losses through the boundaries of the model. Prototype wall temperatures and heat fluxes should be measured to provide better verification of the accuracy of the physical model. The VECTRA computer code is a two-dimensional analytical model. Physical model runs which are isothermal in the Y direction should be made to provide two-dimensional data for more direct comparison to the VECTRA predictions. The ability of the physical model to accurately predict prototype operating conditions should be proven before the model can become a reliable design tool. This will require significantly more prototype operating and glass property data than were available at the time of this study. A complete set of measurements covering power input, heat balances, wall temperatures, glass temperatures, and glass properties should be attempted for at least one prototype run. The information could be used to verify both physical and analytical models. Particle settling and/or sludge buildup should be studied directly by observing the accumulation of the appropriate size and density particles during feeding in the physical model. New designs should be formulated and modeled to minimize the potential problems with melter operation identifed by this study

Development of High-Productivity Continuous Ethanol Production using PVA-Immobilized Zymomonas mobilis in an Immobilized-Cells Fermenter

Directory of Open Access Journals (Sweden)

Nurhayati Nurhayati

2015-07-01

Full Text Available Ethanol as one of renewable energy was being considered an excellent alternative clean-burning fuel to replace gasoline. Continuous ethanol fermentation systems had offered important economic advantages compared to traditional systems. Fermentation rates were significantly improved, especially when continuous fermentation was integrated with cell immobilization techniques to enrich the cells concentration in fermentor. Growing cells of Zymomonas mobilis immobilized in polyvinyl alcohol (PVA gel beads were employed in an immobilized-cells fermentor for continuous ethanol fermentation from glucose. The glucose loading, dilution rate, and cells loading were varied in order to determine which best condition employed in obtaining both high ethanol production and low residual glucose with high dilution rate. In this study, 20 g/L, 100 g/L, 125 g/L and 150 g/L of glucose concentration and 20% (w/v, 40% (w/v and 50% (w/v of cells loading were employed with range of dilution rate at 0.25 to 1 h-1. The most stable production was obtained for 25 days by employing 100 g/L of glucose loading. Meanwhile, the results also exhibited that 125 g/L of glucose loading as well as 40% (w/v of cells loading yielded high ethanol concentration, high ethanol productivity, and acceptable residual glucose at 62.97 g/L, 15.74 g/L/h and 0.16 g/L, respectively. Furthermore, the dilution rate of 4 hour with 100 g/L and 40% (w/v of glucose and cells loading was considered as the optimum condition with ethanol production, ethanol productivity and residual glucose obtained were 49.89 g/L, 12.47 g/L/h, and 2.04 g/L, respectively. This recent study investigated ethanol inhibition as well. The present research had proved that high sugar concentration was successfully converted to ethanol. These achieved results were promising for further study.

One-step synthesis of high-yield biodiesel from waste cooking oils by a novel and highly methanol-tolerant immobilized lipase.

Science.gov (United States)

Wang, Xiumei; Qin, Xiaoli; Li, Daoming; Yang, Bo; Wang, Yonghua

2017-07-01

This study reported a novel immobilized MAS1 lipase from marine Streptomyces sp. strain W007 for synthesizing high-yield biodiesel from waste cooking oils (WCO) with one-step addition of methanol in a solvent-free system. Immobilized MAS1 lipase was selected for the transesterification reactions with one-step addition of methanol due to its much more higher biodiesel yield (89.50%) when compared with the other three commercial immobilized lipases (biodiesel yield (95.45%) was acquired with one-step addition of methanol under the optimized conditions. Moreover, it was observed that immobilized MAS1 lipase retained approximately 70% of its initial activity after being used for four batch cycles. Finally, the obtained biodiesel was further characterized using FT-IR, 1 H and 13 C NMR spectroscopy. These findings indicated that immobilized MAS1 lipase is a promising catalyst for biodiesel production from WCO with one-step addition of methanol under high methanol concentration. Copyright © 2017 Elsevier Ltd. All rights reserved.

Immobilization of Recombinant Glucose Isomerase for Efficient Production of High Fructose Corn Syrup.

Science.gov (United States)

Jin, Li-Qun; Xu, Qi; Liu, Zhi-Qiang; Jia, Dong-Xu; Liao, Cheng-Jun; Chen, De-Shui; Zheng, Yu-Guo

2017-09-01

Glucose isomerase is the important enzyme for the production of high fructose corn syrup (HFCS). One-step production of HFCS containing more than 55% fructose (HFCS-55) is receiving much attention for its industrial applications. In this work, the Escherichia coli harboring glucose isomerase mutant TEGI-W139F/V186T was immobilized for efficient production of HFCS-55. The immobilization conditions were optimized, and the maximum enzyme activity recovery of 92% was obtained. The immobilized glucose isomerase showed higher pH, temperature, and operational stabilities with a K m value of 272 mM and maximum reaction rate of 23.8 mM min -1 . The fructose concentration still retained above 55% after the immobilized glucose isomerase was reused for 10 cycles, and more than 85% of its initial activity was reserved even after 15 recycles of usage at temperature of 90 °C. The results highlighted the immobilized glucose isomerase as a potential biocatalyst for HFCS-55 production.

Four themes that underlie the high-level nuclear waste management program

International Nuclear Information System (INIS)

Sprecher, W.M.

1989-01-01

In 1982, after years of deliberation and in response to mounting pressures from environmental, industrial, and other groups, the US Congress enacted the Nuclear Waste Policy Act (NWPA) of 1982, which was signed into law by the President in January 1983. That legislation signified a major milestone in the nation's management of high-level nuclear waste, since it represented a consensus among the nation's lawmakers to tackle a problem that had evaded solution for decades. Implementation of the NWPA has proven to be exceedingly difficult, as attested by the discord generated by the US Department of Energy's (DOE's) geologic repository and monitored retrievable storage (MRS) facility siting activities. The vision that motivated the crafters of the 1982 act became blurred as opposition to the law increased. After many hearings that underscored the public's concern with the waste management program, the Congress enacted the Nuclear Waste Policy Amendments Act of 1987 (Amendments Act), which steamlined and focused the program, while establishing three independent bodies: the MRS Review Commission, the Nuclear Waste Technical Review Board, and the Office of the Nuclear Waste Negotiator. Yet, even as the program evolves, several themes characterizing the nation's effort to solve the waste management problem continue to prevail. The first of these themes has to do with social consciousness, and the others that follow deal with technical leadership, public involvement and risk perceptions, and program conservatism

Reevaluation Of Vitrified High-Level Waste Form Criteria For Potential Cost Savings At The Defense Waste Processing Facility

International Nuclear Information System (INIS)

Ray, J. W.; Marra, S. L.; Herman, C. C.

2013-01-01

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form

A summary of methods for conditioning and immobilizing ion-exchange resins

International Nuclear Information System (INIS)

Speranzini, R.A.; Buckley, L.P.

1983-02-01

Ion-exchange resins are used in CANDU-PHW nuclear power stations to purify heavy water in the primary heat transport (PHT) and moderator systems. Two promising techniques for conditioning spent ion-exchange resins for disposal have been evaluated: direct immobilization and incineration combined with immobilization of the ash and scrubbed off-gases. When ion-exchange resins were immobilized directly, volumes of bitumen products were about 0.75 times the volumes of untreated resin, while the volumes of cement and polyester products were 2 to 3 times larger. While incinerating the resin is an extra processing step, much smaller volumes result from the latter option. Bitumen and glass product volumes were six and ten times smaller, respectively, than the volumes of untreated resin, while cement and polyester product volumes were about one-half the volume of untreated resin. Since the releases of Cs-136 by leaching were lowest for products made by immobilization in glass, PHT resins which have high concentrations of Cs-137 should be vitrified. Moderator resins which have high concentrations of C-14 should be incinerated and the ash and C-14-contaminated scrubbing solutions should be immobilized. By pretreating such resins with calcium chloride or carbon dioxide, the C-14 present on resin could be released at temperatures below the ignition temperature of the resin. The pretreatment technique reduces the amount of inactive carbon dioxide that must be scrubbed to trap the C-14. The releases of C-14 from immobilized barium hydroxide scrubbing solution were the same as releases from immobilized resin

Enzymatic transesterification of soybean oil with ethanol using lipases immobilized on highly crystalline PVA microspheres

International Nuclear Information System (INIS)

Bergamasco, Juliana; Araujo, Marcelo V. de; Vasconcellos, Adriano de; Luizon Filho, Roberto A.; Hatanaka, Rafael R.; Giotto, Marcus V.; Aranda, Donato A.G.; Nery, José G.

2013-01-01

Polyvinyl alcohol (PVA) microspheres with different degree of crystallinity were used as solid supports for Rhizomucor miehei lipase immobilization, and the enzyme-PVA complexes were used as biocatalysts for the transesterification of soybean oil to fatty acid ethyl esters (FAEE). The amounts of immobilized enzyme on the polymeric supports were similar for both the amorphous microspheres (PVA4) and the high crystalline microspheres (PVA25). However, the enzymatic activity of the immobilized enzymes was depended on the crystallinity degree of the PVA microspheres: enzymes immobilized on the PVA4 microspheres have shown low enzymatic activity (6.13 U mg −1 ), in comparison with enzymes immobilized on the high crystalline PVA25 microspheres (149.15 U mg −1 ). A synergistic effect was observed for the enzyme-PVA25 complex during the transesterification reaction of soybean oil to FAEE: transesterification reactions with free enzyme with the equivalent amount of enzyme that were immobilized onto the PVA25 microspheres (5.4 U) have yielded only 20% of FAEE, reactions with the pure highly crystalline microsphere PVA25 have not yielded FAEE, however reactions with the enzyme-PVA25 complexes have yielded 66.3% of FAEE. This synergistic effect of an immobilized enzyme on a polymeric support has not been observed before for transesterification reaction of triacylglycerides into FAEE. Based on ATR-FTIR, 23 Na- and 13 C-NMR-MAS spectroscopic data and the interaction of the polymeric network intermolecular hydrogen bonds with the lipases residual amino acids a possible explanation for this synergistic effect is provided. Highlights: • Rhizomucor miehei lipase was immobilized on PVA microspheres (PVA4, PVA12, PVA25). • Polymer-enzyme complex was characterized by XDR, SEM, ATR-FTIR, 13 C-CPMAS-NMR, 23 Na-MAS-NMR. • Polymer-enzymes (PVA12 and PVA25) enzymes yielded considerable amount of ethyl esters. • Synergistic effect was observed for the polymer-enzyme complexes

A comparison of high-level waste form characteristics

International Nuclear Information System (INIS)

Salmon, R.; Notz, K.J.

1991-01-01

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

Tentative job analysis for a high-level, fixed-site, nuclear security officer

International Nuclear Information System (INIS)

Adams, K.G.; Trujillo, A.A.

1977-10-01

A tentative job analysis for a high-level, fixed-site, nuclear security officer is presented. The primary objective of the report is to provide a framework for evaluating the functions of a security officer in physical protection systems. Several job requirements related to duties, basic skills, personal contacts, supervision, working conditions, and decision making are presented. Individual character traits desirable in security officers are described

Magnesium Potassium Phosphate Compound for Immobilization of Radioactive Waste Containing Actinide and Rare Earth Elements

Directory of Open Access Journals (Sweden)

Sergey E. Vinokurov

2018-06-01

Full Text Available The problem of effective immobilization of liquid radioactive waste (LRW is key to the successful development of nuclear energy. The possibility of using the magnesium potassium phosphate (MKP compound for LRW immobilization on the example of nitric acid solutions containing actinides and rare earth elements (REE, including high level waste (HLW surrogate solution, is considered in the research work. Under the study of phase composition and structure of the MKP compounds that is obtained by the XRD and SEM methods, it was established that the compounds are composed of crystalline phases—analogues of natural phosphate minerals (struvite, metaankoleite. The hydrolytic stability of the compounds was determined according to the semi-dynamic test GOST R 52126-2003. Low leaching rates of radionuclides from the compound are established, including a differential leaching rate of 239Pu and 241Am—3.5 × 10−7 and 5.3 × 10−7 g/(cm2∙day. As a result of the research work, it was concluded that the MKP compound is promising for LRW immobilization and can become an alternative material combining the advantages of easy implementation of the technology, like cementation and the high physical and chemical stability corresponding to a glass-like compound.

Laboratory characterization and vitrification of Hanford radioactive high-level waste

International Nuclear Information System (INIS)

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

1991-05-01

Radioactive high-level wastes generated at the Department of Energy's Hanford Site are stored in underground carbon steel tanks. Two double-shell tanks contain neutralized current acid waste (NCAW) from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant. The tanks were sampled for characterization and waste immobilization process/product development. The high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). The CAW was ''neutralized'' to a pH of approximately 14 by adding sodium hydroxide to reduce corrosion of the tanks. This ''neutralized'' waste is called Neutralized Current Acid Waste. Both precipitated solids and liquids are stored in the NCAW waste tanks. The NCAW contains small amounts of plutonium and most of the fission products and americium from the irradiated fuel. NCAW also contains stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. The NCAW will be retrieved, pretreated, and immobilized prior to final disposal. Pretreatment consists of water washing the precipitated NCAW solids for sulfate and soluble salts removal as a waste reduction step prior to vitrification. This waste is expected to be the first waste type to be retrieved and vitrified in the Hanford Waste Vitrification Plant (HWVP). A characterization plan was developed that details the processing of the small-volume NCAW samples through retrieval, pretreatment and vitrification process steps. Physical, rheological, chemical, and radiochemical properties were measured throughout these process steps. The results of nonradioactive simulant tests were used to develop appropriate pretreatment and vitrification process steps. The processing and characterization of simulants and actual NCAW tank samples are used to evaluate the operation of these processes. 3 refs., 1 fig., 4 tabs

Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Murphy, W.M.; Kovach, L.A.

1995-01-01

A workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste (HLW) was held in San Antonio, Texas, on July 22-25, 1991. It was sponsored by the US Nuclear Regulatory Commission (NRC) and the Center for Nuclear Waste Regulatory Analyses (CNWRA). Invitations to the workshop were extended to a large number of individuals with a variety of technical and professional interests related to geologic disposal of nuclear waste and natural analog studies. The objective of the workshop was to examine the role of natural analog studies in performance assessment, site characterization, and prioritization of research related to geologic disposal of HLW

Nuclear Level Densities

International Nuclear Information System (INIS)

Grimes, S.M.

2005-01-01

Recent research in the area of nuclear level densities is reviewed. The current interest in nuclear astrophysics and in structure of nuclei off of the line of stability has led to the development of radioactive beam facilities with larger machines currently being planned. Nuclear level densities for the systems used to produce the radioactive beams influence substantially the production rates of these beams. The modification of level-density parameters near the drip lines would also affect nucleosynthesis rates and abundances

A High-Resolution Magic Angle Spinning NMR Study of the Enantiodiscrimination of 3,4-Methylenedioxymethamphetamine (MDMA by an Immobilized Polysaccharide-Based Chiral Phase.

Directory of Open Access Journals (Sweden)

Juliana C Barreiro

Full Text Available This paper reports the investigation of the chiral interaction between 3,4-methylenedioxy-methamphetamine (MDMA enantiomers and an immobilized polysaccharide-based chiral phase. For that, suspended-state high-resolution magic angle spinning nuclear magnetic resonance spectroscopy (1H HR-MAS NMR was used. 1H HR-MAS longitudinal relaxation time and Saturation Transfer Difference (STD NMR titration experiments were carried out yielding information at the molecular level of the transient diastereoisomeric complexes of MDMA enantiomers and the chiral stationary phase. The interaction of the enantiomers takes place through the aromatic moiety of MDMA and the aromatic group of the chiral selector by π-π stacking for both enantiomers; however, a stronger interaction was observed for the (R-enantiomer, which is the second one to elute at the chromatographic conditions.

High-level waste-form-product performance evaluation

International Nuclear Information System (INIS)

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

1982-01-01

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

Immobilization of hazardous and radioactive waste into glass structures

International Nuclear Information System (INIS)

Wicks, G.G.

1997-01-01

As a result of more than three decades of international research, glass has emerged as the material of choice for immobilization of a wide range of potentially hazardous radioactive and non-radioactive materials. The ability of glass structures to incorporate and then immobilize many different elements into durable, high integrity, waste glass products is a direct function of the unique random network structure of the glassy state. Every major country involved with long-term management of high-level radioactive waste (HLW) has either selected or is considering glass as the matrix of choice for immobilizing and ultimately, disposing of the potentially hazardous, high-level radioactive material. There are many reasons why glass is preferred. Among the most important considerations are the ability of glass structures to accommodate and immobilize the many different types of radionuclides present in HLW, and to produce a product that not only has excellent technical properties, but also possesses good processing features. Good processability allows the glass to be fabricated with relative ease even under difficult remote-handling conditions necessary for vitrification of highly radioactive material. The single most important property of the waste glass produced is its ability to retain hazardous species within the glass structure and this is reflected by its excellent chemical durability and corrosion resistance to a wide range of environmental conditions. In addition to immobilization of HLW glass matrices are also being considered for isolation of many other types of hazardous materials, both radioactive as well as nonradioactive. This includes vitrification of various actinides resulting from clean-up operations and the legacy of the cold war, as well as possible immobilization of weapons grade plutonium resulting from disarmament activities. Other types of wastes being considered for immobilization into glasses include transuranic wastes, mixed wastes, contaminated

Immobilization of biomolecules on cysteamine-modified polyaniline film for highly sensitive biosensing.

Science.gov (United States)

Cai, Qi; Xu, Baojian; Ye, Lin; Di, Zengfeng; Zhang, Jishen; Jin, Qinghui; Zhao, Jianlong; Xue, Jian; Chen, Xianfeng

2014-03-01

We present a new cysteamine (CS)-modified polyaniline (PANI) film for highly efficient immobilization of biomolecules in biosensing technology. This electrochemical deposited PANI film treated with CS and glutaraldehyde could be employed as an excellent substrate for biomolecules immobilization. The parameters of PANI growth were optimized to obtain suitable surface morphology of films for biomolecules combination with the help of electron and atomic force microscopy. Cyclic voltammetry (CV) was utilized to illustrate the different electrochemical activities of each modified electrode. Due to the existence of sulfydryl group and amino group in CS, surface modification with CS was proven to reduce oxidized units on PANI film remarkably, as evidenced by both ATR-FTIR and Raman spectroscopy characterizations. Furthermore, bovine serum albumin (BSA) was used as the model protein to investigate the immobilization efficiency of biomolecules on the PANI film, comparative study using quartz crystal microbalance (QCM) showed that BSA immobilized on CS-modified PANI could be increased by at least 20% than that without CS-modified PANI in BSA solution with the concentration of 0.1-1mg/mL. The CS-modified PANI film would be significant for the immobilization and detection of biomolecules and especially promising in the application of immunosensor for ultrasensitive detection. © 2013 Published by Elsevier B.V.

The nuclear physical method for high pressure steam manifold water level gauging and its error

International Nuclear Information System (INIS)

Li Nianzu; Li Beicheng; Jia Shengming

1993-10-01

A new method, which is non-contact on measured water level, for measuring high pressure steam manifold water level with nuclear detection technique is introduced. This method overcomes the inherent drawback of previous water level gauges based on other principles. This method can realize full range real time monitoring on the continuous water level of high pressure steam manifold from the start to full load of boiler, and the actual value of water level can be obtained. The measuring errors were analysed on site. Errors from practical operation in Tianjin Junliangcheng Power Plant and in laboratory are also presented

Production of Structured Phosphatidylcholine with High Content of DHA/EPA by Immobilized Phospholipase A1-Catalyzed Transesterification

Directory of Open Access Journals (Sweden)

Xiang Li

2014-08-01

Full Text Available This paper presents the synthesis of structured phosphatidylcholine (PC enriched with docosahexaenoic acid (DHA and eicosapentaenoic acid (EPA by transesterification of DHA/EPA-rich ethyl esters with PC using immobilized phospholipsase A1 (PLA1 in solvent-free medium. Firstly, liquid PLA1 was immobilized on resin D380, and it was found that a pH of 5 and a support/PLA1 ratio (w/v of 1:3 were the best conditions for the adsorption. Secondly, the immobilized PLA1 was used to catalyze transesterification of PC and DHA/EPA-rich ethyl esters. The maximal incorporation of DHA and EPA achieved was 30.7% for 24 h of reaction at 55 °C using a substrate mass ratio (PC/ethyl esters of 1:6, an immobilized PLA1 loading of 15% and water dosage of 1.25%. Then the reaction mixture was analyzed by 31P nuclear magnetic resonance (NMR. The composition of reaction product included 16.5% PC, 26.3% 2-diacyl-sn-glycero-3-lysophosphatidylcholine (1-LPC, 31.4% 1-diacyl-sn-glycero-3-lysophosphatidylcholine (2-LPC, and 25.8% sn-glycerol-3-phosphatidylcholine (GPC.

Canada's high-level nuclear waste disposal concept and its evaluation process

International Nuclear Information System (INIS)

Sheng, Grant; Shemilt, L.W.

2004-01-01

The concept of disposing high-level nuclear waste in granitic rocks in the Canadian Shield, developed by Atomic Energy of Canada Limited (AECL), is anticipated to undergo a national public review within two years. The disposal concept, its documentation, and its process of evaluation, including the role of the public, government and the scientific/engineering community, are summarized. A Technical Advisory Committee (TAC) has provided external peer review of the Program since 1979 and its findings are published in annual reports which are publicly available. (author)

High production of D-tagatose, a potential sugar substitute, using immobilized L-arabinose isomerase.

Science.gov (United States)

Kim, P; Yoon, S H; Roh, H J; Choi, J H

2001-01-01

An L-arabinose isomerase of Escherichia coli was immobilized using covalent binding to agarose to produce D-tagatose, a bulking sweetener that can be economically used as a sugar substitute. The immobilized L-arabinose isomerase stably produced an average of 7.5 g-tagatose/L.day for 7 days with a productivity exceeding that of the free enzyme (0.47 vs 0.30 mg/U.day). Using a scaled-up immobilized enzyme system, 99.9 g-tagatose/L was produced from galactose with 20% equilibrium in 48 h. The process was repeated two more times with production of 104.1 and 103.5 g-tagatose/L. D-Tagatose production using an immobilized L-arabinose isomerase has a high potential for commercial application.

Glass science tutorial: Lecture No. 8, introduction cementitious systems for Low-Level Waste immobilization

International Nuclear Information System (INIS)

Young, J.F.; Kirkpatrick, R.J.; Mason, T.O.; Brough, A.

1995-07-01

This report presents details about cementitious systems for low-level waste immobilization. Topics discussed include: composition and properties of portland cement; hydration properties; microstructure of concrete; pozzolans; slags; zeolites; transport properties; and geological aspects of long-term durability of concrete

Glass science tutorial: Lecture No. 8, introduction cementitious systems for Low-Level Waste immobilization

Energy Technology Data Exchange (ETDEWEB)

Young, J.F.; Kirkpatrick, R.J.; Mason, T.O.; Brough, A.

1995-07-01

This report presents details about cementitious systems for low-level waste immobilization. Topics discussed include: composition and properties of portland cement; hydration properties; microstructure of concrete; pozzolans; slags; zeolites; transport properties; and geological aspects of long-term durability of concrete.

Permitting plan for the high-level waste interim storage

International Nuclear Information System (INIS)

Deffenbaugh, M.L.

1997-01-01

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

Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

Science.gov (United States)

Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

2005-12-01

Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

Review of radiation effects in solid-nuclear-waste forms

International Nuclear Information System (INIS)

Weber, W.J.

1981-09-01

Radiation effects on the stability of high-level nuclear waste (HLW) forms are an important consideration in the development of technology to immobilize high-level radioactive waste because such effects may significantly affect the containment of the radioactive waste. Since the required containment times are long (10 3 to 10 6 years), an understanding of the long-term cumulative effects of radiation damage on the waste forms is essential. Radiation damage of nuclear waste forms can result in changes in volume, leach rate, stored energy, structure/microstructure, and mechanical properties. Any one or combination of these changes might significantly affect the long-term stability of the nuclear waste forms. This report defines the general radiation damage problem in nuclear waste forms, describes the simulation techniques currently available for accelerated testing of nuclear waste forms, and reviews the available data on radiation effects in both glass and ceramic (primarily crystalline) waste forms. 76 references

The socioeconomic impacts of high-level nuclear waste facility siting

International Nuclear Information System (INIS)

Murdock, S.H.; Leistritz, F.L.; Hamm, R.R.

1987-01-01

High-level nuclear waste repositories will be located in sparsely settled rural areas in the U.S. These projects will significantly effect the economic, demographic, public service, fiscal, and social (the socioeconomic) dimensions of those rural areas. This paper examines some of the potential socioeconomic impacts and the characteristics of mitigation programs necessary, if these impacts are to be addressed. Both standard impacts, those resulting from the fact that--like many other large-scale developments--repositories will involve a substantial number of new workers and residents (relative to the size of existing communities) and special impacts, those resulting from the fact that repositories store radioactive materials, are examined

Study of powellite-rich glass-ceramics for nuclear waste immobilization

International Nuclear Information System (INIS)

Taurines, T.

2012-01-01

MoO 3 is poorly soluble in borosilicate glasses which can lead to the crystallization of undesired phases when its concentration or the charge load (minor actinides and fission products concentration) is too high. Crystallization control is needed to guarantee good immobilization properties. We studied powellite-rich glass-ceramics obtained from a simplified nuclear glass in the system SiO 2 - B 2 O 3 - Na 2 O - CaO - Al 2 O 3 - MoO 3 - RE 2 O 3 (RE = Gd, Eu, Nd) by various heat treatments. Rare earth elements (REE) were added as minor actinides surrogates and as spectroscopic probes. The influence of MoO 3 and RE 2 O 3 content on powellite (CaMoO 4 ) crystallization was investigated. Various glass-ceramics (similar residual glass + powellite) were obtained with large crystal size distributions. Phase separation due to molybdenum occurs during quenching when [MoO 3 ] ≥ 2.5 mol%. We showed that increasing the rare earth content can suppress the phase separation due to molybdenum but it leads to spinodal decomposition of the residual glass. Furthermore, we studied the effects of parent glass complexifying and the insertion of Gd 3+ ions into the powellite structure. In order to understand the influence of microstructure on evolutions under β-irradiation, we studied point defects creation and structural changes. We showed that the damage induced by electronic excitations in the glass-ceramics is driven by the damage in the residual glass. (author) [fr

Changes in angiotensin AT1 receptor mRNA levels in the rat brain after immobilization stress and inhibition of central nitric oxide synthase.

Science.gov (United States)

Kiss, A; Jurkovicova, D; Jezova, D; Krizanova, O

2001-06-01

To study functional interactions between angiotensin II AT1 receptors and nitric oxide (NO) activity in different brain areas in rats exposed to immobilization stress. Central inhibition of nitric oxide synthase (NOS) was provided by intracerebroventricular (i.c.v.) administration of (N-omega-nitro-L-arginine-methylester) L-NAME and analysis of AT1 receptor mRNA was performed using semiquantitative reverse transcription-polymerase chain reaction (RT-PCR) technique. The immobilization in prone position lasted 2 hrs and the rats were sacrificed 24 hr later. The hypothalamus, hippocampus, thalamus, and cortex were isolated from fresh brains. In the cortex, gene expression of AT1 receptors was unaffected either by L-NAME treatment, or by a single exposure to immobilization stress for 2 hours followed by 24 hours of rest. In the hippocampus, the repeated treatment with L-NAME increased mRNA levels of AT1 receptors approximately 9-times compared to those in the control (untreated) group. Immobilization also increased AT1 receptor mRNA levels in the hippocampus which was similar to that induced by the L-NAME. The increase of AT1 receptor mRNA levels in the hippocampus of immobilized rats was not further altered when the animals were pretreated with L-NAME. In control rats, exposure to immobilization resulted in a significant rise in mRNA levels coding for AT1 receptors in the hypothalamus, but not in the thalamus. L-NAME treatment showed a tendency of increase in AT1 receptor mRNA levels in the hypothalamus. Moreover, when animals treated with L-NAME were subjected to immobilization, a further increase in AT1 receptor mRNA levels was observed in the hypothalamus in comparison with corresponding controls. The present data indicate that a single immobilization stress results in increased gene expression of AT1 receptors in the hypothalamus and hippocampus. The rise in AT1 mRNA levels in the same brain structures after repeated treatment with L-NAME allow to suggest an

Plutonium disposition via immobilization in ceramic or glass

Energy Technology Data Exchange (ETDEWEB)

Gray, L.W.; Kan, T.; Shaw, H.F.; Armantrout, A.

1997-03-05

The management of surplus weapons plutonium is an important and urgent task with profound environmental, national, and international security implications. In the aftermath of the Cold War, Presidential Policy Directive 13, and various analyses by renown scientific, technical, and international policy organizations have brought about a focused effort within the Department of Energy to identify and implement paths for the long term disposition of surplus weapons- usable plutonium. The central goal of this effort is to render surplus weapons plutonium as inaccessible and unattractive for reuse in nuclear weapons as the much larger and growing stock of plutonium contained in spent fuel from civilian reactors. One disposition option being considered for surplus plutonium is immobilization, in which the plutonium would be incorporated into a glass or ceramic material that would ultimately be entombed permanently in a geologic repository for high-level waste.

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

Energy Technology Data Exchange (ETDEWEB)

Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

Energy Technology Data Exchange (ETDEWEB)

Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

High performance gamma measurements of equipment retrieved from Hanford high-level nuclear waste tanks

International Nuclear Information System (INIS)

Troyer, G.L.

1997-01-01

The cleanup of high level defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to 90% saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed

High performance gamma measurements of equipment retrieved from Hanford high-level nuclear waste tanks

Energy Technology Data Exchange (ETDEWEB)

Troyer, G.L.

1997-03-17

The cleanup of high level defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to 90% saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed.

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

Energy Technology Data Exchange (ETDEWEB)

Rusin, J.M.

1980-12-01

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

Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

International Nuclear Information System (INIS)

McDonell, W.R.

1982-04-01

An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

Immobilization of acid digestion residue

International Nuclear Information System (INIS)

Greenhalgh, W.O.; Allen, C.R.

1983-01-01

Acid digestion treatment of nuclear waste is similar to incineration processes and results in the bulk of the waste being reduced in volume and weight to some residual solids termed residue. The residue is composed of various dispersible solid materials and typically contains the resultant radioactivity from the waste. This report describes the immobilization of the residue in portland cement, borosilicate glass, and some other waste forms. Diagrams showing the cement and glass virtification parameters are included in the report as well as process steps and candidate waste product forms. Cement immobilization is simplest and probably least expensive; glass vitrification exhibits the best overall volume reduction ratio

Mathematical modelling of heat production in deep geological repository of high-level nuclear waste

International Nuclear Information System (INIS)

Kovanda, O.

2017-01-01

Waste produced by nuclear industry requires special handling. Currently, there is a research taking place, focused at possibilities of nuclear waste storage in deep geological repositories, hosted in stable geological environment. The high-level nuclear waste produces significant amount of heat for a long time, which can affect either environment outside of or within the repository in a negative way. Therefore to reduce risks, it is desirable to know the principles of such heat production, which can be achieved using mathematical modeling. This thesis comes up with a general model of heat production-time dependency, dependable on initial composition of the waste. To be able to model real situations, output of this thesis needs to be utilized in an IT solution. (authors)

Reevaluation of Vitrified High-Level Waste Form Criteria for Potential Cost Savings at the Defense Waste Processing Facility - 13598

Energy Technology Data Exchange (ETDEWEB)

Ray, J.W. [Savannah River Remediation (United States); Marra, S.L.; Herman, C.C. [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

2013-07-01

At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

Sorption of strontium on uranyl peroxide: implications for a high-level nuclear waste repository.

Science.gov (United States)

Sureda, Rosa; Martínez-Lladó, Xavier; Rovira, Miquel; de Pablo, Joan; Casas, Ignasi; Giménez, Javier

2010-09-15

Strontium-90 is considered the most important radioactive isotope in the environment and one of the most frequently occurring radionuclides in groundwaters at nuclear facilities. The uranyl peroxide studtite (UO2O2 . 4H2O) has been observed to be formed in spent nuclear fuel leaching experiments and seems to have a relatively high sorption capacity for some radionuclides. In this work, the sorption of strontium onto studtite is studied as a function of time, strontium concentration in solution and pH. The main results obtained are (a) sorption is relatively fast although slower than for cesium; (b) strontium seems to be sorbed via a monolayer coverage of the studtite surface, (c) sorption has a strong dependence on ionic strength, is negligible at acidic pH, and increases at neutral to alkaline pH (almost 100% of the strontium in solution is sorbed above pH 10). These results point to uranium secondary solid phase formation on the spent nuclear fuel as an important mechanism for strontium retention in a high-level nuclear waste repository (HLNW). Copyright 2010 Elsevier B.V. All rights reserved.

Advanced BorobondTM Shields for Nuclear Materials Containment and BorobondTM Immobilization of Volatile Fission Products - Final CRADA Report

International Nuclear Information System (INIS)

Wagh, Arun S.

2016-01-01

Borobond is a company-proprietary material developed by the CRADA partner in collaboration with Argonne, and is based on Argonne's Ceramicrete technology. It is being used by DOE for nuclear materials safe storage, and Boron Products, LLC is the manufacturer and supplier of Borobond. The major objective of this project was to produce a more versatile composition of this material and find new applications. Major target applications were use for nuclear radiation shields, such as in dry storage casks; use in immobilization of most difficult waste streams, such as Hanford K-Basin waste; use for soluble and volatile fission products, such as Cs, Tc, Sr, and I; and use for corrosion and fire protection applications in nuclear facilities.

Instrument reliability for high-level nuclear-waste-repository applications

International Nuclear Information System (INIS)

Rogue, F.; Binnall, E.P.; Armantrout, G.A.

1983-01-01

Reliable instrumentation will be needed to evaluate the characteristics of proposed high-level nuclear-wasted-repository sites and to monitor the performance of selected sites during the operational period and into repository closure. A study has been done to assess the reliability of instruments used in Department of Energy (DOE) waste repository related experiments and in other similar geological applications. The study included experiences with geotechnical, hydrological, geochemical, environmental, and radiological instrumentation and associated data acquisition equipment. Though this paper includes some findings on the reliability of instruments in each of these categories, the emphasis is on experiences with geotechnical instrumentation in hostile repository-type environments. We review the failure modes, rates, and mechanisms, along with manufacturers modifications and design changes to enhance and improve instrument performance; and include recommendations on areas where further improvements are needed

Application of spent fuel treatment technology to plutonium immobilization

International Nuclear Information System (INIS)

McPheeters, C.C.; Ackerman, J.P.; Gay, E.C., Johnson, G.K.

1996-01-01

The purpose of the electrometallurgical treatment technology being developed at Argonne National Laboratory (ANL) is to convert certain spent nuclear fuels into waste forms that are suitable for disposal in a geological repository for nuclear waste. The spent fuels of interest are those that cannot be safely stored for a long time in their current condition, and those that cannot be qualified for repository disposal. This paper explores the possibility of applying this electrometallurgical treatment technology to immobilization of surplus fissile materials, primarily plutonium. Immobilization of surplus fissile materials by electrometallurgical treatment could be done in the same facilities, at the same time. and in the same equipment as the proposed treatment of the present inventory of spent nuclear fuel. The cost and schedule savings of this simultaneous treatment scheme would be significant

Hanford immobilized LAW product acceptance: Initial Tanks Focus Area testing data package

Energy Technology Data Exchange (ETDEWEB)

JD Vienna; A Jiricka; BP McGrail; BM Jorgensen; DE Smith; BR Allen; JC Marra; DK Peeler; KG Brown; IA Reamer; WL Ebert

2000-03-08

The Hanford Site's mission has been to produce nuclear materials for the US Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during plutonium production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The total volume of LAW requiring immobilization will include the LAW separated from the tank waste, as well as new wastes generated by the retrieval, pretreatment, and immobilization processes. Per the Tri-Party Agreement (1994), both the LAW and HLW will be vitrified. It has been estimated that vitrification of the LAW waste will result in over 500,000 metric tons or 200,000 m{sup 3} of immobilized LAW (ILAW) glass. The ILAW glass is to be disposed of onsite in a near-surface burial facility. It must be demonstrated that the disposal system will adequately retain the radionuclides and prevent contamination of the surrounding environment. This report describes a study of the impacts of systematic glass-composition variation on the responses from accelerated laboratory corrosion tests of representative LAW glasses. A combination of two tests, the product consistency test and vapor-hydration test, is being used to give indictations of the relative rate at which a glass could be expected to corrode in the burial scenario.

Hanford immobilized LAW product acceptance: Initial Tanks Focus Area testing data package

International Nuclear Information System (INIS)

JD Vienna; A Jiricka; BP McGrail; BM Jorgensen; DE Smith; BR Allen; JC Marra; DK Peeler; KG Brown; IA Reamer; WL Ebert

2000-01-01

The Hanford Site's mission has been to produce nuclear materials for the US Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during plutonium production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The total volume of LAW requiring immobilization will include the LAW separated from the tank waste, as well as new wastes generated by the retrieval, pretreatment, and immobilization processes. Per the Tri-Party Agreement (1994), both the LAW and HLW will be vitrified. It has been estimated that vitrification of the LAW waste will result in over 500,000 metric tons or 200,000 m 3 of immobilized LAW (ILAW) glass. The ILAW glass is to be disposed of onsite in a near-surface burial facility. It must be demonstrated that the disposal system will adequately retain the radionuclides and prevent contamination of the surrounding environment. This report describes a study of the impacts of systematic glass-composition variation on the responses from accelerated laboratory corrosion tests of representative LAW glasses. A combination of two tests, the product consistency test and vapor-hydration test, is being used to give indictations of the relative rate at which a glass could be expected to corrode in the burial scenario

United States Program on Spent Nuclear Fuel and High-Level Radioactive Waste Management

International Nuclear Information System (INIS)

Stewart, L.

2004-01-01

The President signed the Congressional Joint Resolution on July 23, 2002, that designated the Yucca Mountain site for a proposed geologic repository to dispose of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The United States (U.S.) Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently focusing its efforts on submitting a license application to the U.S. Nuclear Regulatory Commission (NRC) in December 2004 for construction of the proposed repository. The legislative framework underpinning the U.S. repository program is the basis for its continuity and success. The repository development program has significantly benefited from international collaborations with other nations in the Americas

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Plutonium immobilization plant using ceramic in existing facilities at the Savannah River site

International Nuclear Information System (INIS)

DiSabatino, A.

1998-01-01

The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources, and through a ceramic immobilization process converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors. The ceramic immobilization alternative presented in this report consists of first converting the surplus material to an oxide, followed by incorporating the plutonium oxide into a titanate-based ceramic material that is placed in metal cans

Radiation Synthesis of Nanogel for Bioactives Immobilization

International Nuclear Information System (INIS)

Hamzah, M.Y.

2009-01-01

Both hydrophilic and hydrophobic core nanogel are currently being developed for immobilization and delivery purposes in Malaysian Nuclear Agency. Hydrophilic nanogel is produced by using inverse micelles irradiation of polyethelyne glycol diacrylate (PEGDA). The hydrophobic nanogel is produced via irradiation of acrylated form of palm oil. These nanogels will be used to immobilize bio actives such as curcumin, tyhmoquinone, oryzanol and chitosan. Preliminary investigation of the nanogel size using dynamic light scattering (DLS) shows that nanogel with sizes below 100nm can be obtained. (author)

Radiation Synthesis of Nanogel for Bioactives Immobilization

Energy Technology Data Exchange (ETDEWEB)

Hamzah, M. Y. [Polymer Modification Group, Malaysian Nuclear Agency, Bangi (Malaysia)

2009-07-01

Both hydrophilic and hydrophobic core nanogel are currently being developed for immobilization and delivery purposes in Malaysian Nuclear Agency. Hydrophilic nanogel is produced by using inverse micelles irradiation of polyethelyne glycol diacrylate (PEGDA). The hydrophobic nanogel is produced via irradiation of acrylated form of palm oil. These nanogels will be used to immobilize bio actives such as curcumin, tyhmoquinone, oryzanol and chitosan. Preliminary investigation of the nanogel size using dynamic light scattering (DLS) shows that nanogel with sizes below 100nm can be obtained. (author)

High-level waste processing and disposal

International Nuclear Information System (INIS)

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

1984-11-01

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

Selection of a glass-ceramic formulation to immobilize fluorinel- sodium calcine

International Nuclear Information System (INIS)

Staples, B.A.; Wood, H.C.

1994-12-01

One option for immobilizing calcined high level wastes produced by nuclear fuel reprocessing activities at the Idaho Chemical Processing Plant (ICPP) is conversion to a glass-ceramic form through hot isostatic pressing. Calcines exist in several different chemical compositions, and thus candidate formulations have been developed for converting each to glass-ceramic forms which are potentially resistant to aqueous corrosion and stable enough to qualify for repository storage. Fluorinel/Na, a chemically complex calcine type, is one of the types being stored at ICPP, and development efforts have identified three formulations with potential for immobilizing it. These are a glass forming additive that uses aluminum metal to enhance reactivity, a second glass forming additive that uses titanium metal to enhance reactivity and a third that uses not only a combination of silicon and titanium metals but enough phosphorous pentoxide to form a calcium phosphate host phase in the glass-ceramic product. Glass-ceramics of each formulation performed well in restricted characterization tests. However, none of the three was subjected to rigorous testing that would provide information on whether each was processable, that is able to retain favorable characteristics over a practical range of processing conditions

High density gold nanoparticles immobilized on surface via plasma deposited APTES film for decomposing organic compounds in microchannels

Science.gov (United States)

Rao, Xi; Guyon, Cédric; Ognier, Stephanie; Da Silva, Bradley; Chu, Chenglin; Tatoulian, Michaël; Hassan, Ali Abou

2018-05-01

Immobilization of colloidal particles (e.g. gold nanoparticles (AuNps)) on the inner surface of micro-/nano- channels has received a great interest for catalysis. A novel catalytic ozonation setup using a gold-immobilized microchannel reactor was developed in this work. To anchor AuNps, (3-aminopropyl) triethoxysilane (APTES) with functional amine groups was deposited using plasma enhanced chemical vapor deposition (PECVD) process. The results clearly evidenced that PECVD processing exhibited relatively high efficiency for grafting amine groups and further immobilizing AuNPs. The catalytic activity of gold immobilized microchannel was evaluated by pyruvic acid ozonation. The decomposition rate calculated from High Performance Liquid Chromatography (HPLC) indicated a much better catalytic performance of gold in microchannel than that in batch. The results confirmed immobilizing gold nanoparticles on plasma deposited APTES for preparing catalytic microreactors is promising for the wastewater treatment in the future.

Combination of microbial oxidation and biogenic schwertmannite immobilization: A potential remediation for highly arsenic-contaminated soil.

Science.gov (United States)

Yang, Zhihui; Wu, Zijian; Liao, Yingping; Liao, Qi; Yang, Weichun; Chai, Liyuan

2017-08-01

Here, a novel strategy that combines microbial oxidation by As(III)-oxidizing bacterium and biogenic schwertmannite (Bio-SCH) immobilization was first proposed and applied for treating the highly arsenic-contaminated soil. Brevibacterium sp. YZ-1 isolated from a highly As-contaminated soil was used to oxidize As(III) in contaminated soils. Under optimum culture condition for microbial oxidation, 92.3% of water-soluble As(III) and 84.4% of NaHCO 3 -extractable As(III) in soils were removed. Bio-SCH synthesized through the oxidation of ferrous sulfate by Acidithiobacillus ferrooxidans immobilize As(V) in the contaminated soil effectively. Consequently, the combination of microbial oxidation and Bio-SCH immobilization performed better in treating the highly As-contaminated soil with immobilization efficiencies of 99.3% and 82.6% for water-soluble and NaHCO 3 -extractable total As, respectively. Thus, the combination can be considered as a green remediation strategy for developing a novel and valuable solution for As-contaminated soils. Copyright © 2017 Elsevier Ltd. All rights reserved.

Alginate Immobilization of Metabolic Enzymes (AIME) for High-Throughput Screening Assays (SOT)

Science.gov (United States)

Alginate Immobilization of Metabolic Enzymes (AIME) for High-Throughput Screening Assays DE DeGroot, RS Thomas, and SO SimmonsNational Center for Computational Toxicology, US EPA, Research Triangle Park, NC USAThe EPA’s ToxCast program utilizes a wide variety of high-throughput s...

Immobilized enzymes and cells

Energy Technology Data Exchange (ETDEWEB)

Bucke, C; Wiseman, A

1981-04-04

This article reviews the current state of the art of enzyme and cell immobilization and suggests advances which might be made during the 1980's. Current uses of immobilized enzymes include the use of glucoamylase in the production of glucose syrups from starch and glucose isomerase in the production of high fructose corn syrup. Possibilities for future uses of immobilized enzymes and cells include the utilization of whey and the production of ethanol.

Why is the South Orkney Island shelf (the world's first high seas marine protected area) a carbon immobilization hotspot?

Science.gov (United States)

Barnes, David K A; Ireland, Louise; Hogg, Oliver T; Morley, Simon; Enderlein, Peter; Sands, Chester J

2016-03-01

The Southern Ocean archipelago, the South Orkney Islands (SOI), became the world's first entirely high seas marine protected area (MPA) in 2010. The SOI continental shelf (~44 000 km(2) ), was less than half covered by grounded ice sheet during glaciations, is biologically rich and a key area of both sea surface warming and sea-ice losses. Little was known of the carbon cycle there, but recent work showed it was a very important site of carbon immobilization (net annual carbon accumulation) by benthos, one of the few demonstrable negative feedbacks to climate change. Carbon immobilization by SOI bryozoans was higher, per species, unit area and ice-free day, than anywhere-else polar. Here, we investigate why carbon immobilization has been so high at SOI, and whether this is due to high density, longevity or high annual production in six study species of bryozoans (benthic suspension feeders). We compared benthic carbon immobilization across major regions around West Antarctica with sea-ice and primary production, from remotely sensed and directly sampled sources. Lowest carbon immobilization was at the northernmost study regions (South Georgia) and southernmost Amundsen Sea. However, data standardized for age and density showed that only SOI was anomalous (high). High immobilization at SOI was due to very high annual production of bryozoans (rather than high densities or longevity), which were 2x, 3x and 5x higher than on the Bellingshausen, South Georgia and Amundsen shelves, respectively. We found that carbon immobilization correlated to the duration (but not peak or integrated biomass) of phytoplankton blooms, both in directly sampled, local scale data and across regions using remote-sensed data. The long bloom at SOI seems to drive considerable carbon immobilization, but sea-ice losses across West Antarctica mean that significant carbon sinks and negative feedbacks to climate change could also develop in the Bellingshausen and Amundsen seas. © 2015 John Wiley

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

International Nuclear Information System (INIS)

Mann, S.A.

1981-01-01

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

A rapid and continuous system for immobilization of nuclear material waste by vitrification

International Nuclear Information System (INIS)

Shareef, M.U.; Hussain, S.H.; Tufail, M.; Rashid, F.

2009-01-01

The nuclear technology has prime importance and backbone for rapid development of medical sciences, industries and in power generation as an alternate source of energy. Despite all these facts there is a major problem which is always associated with nuclear technology, that is, the generation of undesirable radioactive wastes. The radioactive wastes are quite problematic and need major attention for its treatment, conditioning and properly disposal to keep the environmental activities and human ecosystem healthy and safe. There are different large scale methods and processes to treat and dispose off the radioactive wastes. These processes are evaluated and designed by the various world competent and pronounced scientists in the light of rules and safety limits set by IAEA and other regulatory authorities to protect the environment and eventually protect our ecosystem. The research and development work on radioactive waste has been proceeding for the last fifty years but still it is a core issue and a big challenge for the nuclear scientists and radiation workers. In this study a rapid and continuous system for immobilization of nuclear waste into glass matrix by vitrification has been designed. In general treatment methods, Borosilicate glass is preferred because it is efficient, cost effective and rapid to that of other radioactive waste form. In this process the simulated waste is mixed with glass forming material and process for melting to form a glassy substrate in continuous manner. The waste is being converted into vitreous form and encapsulate into a glass matrix. (author)

Permitting plan for the immobilized low-activity waste project

International Nuclear Information System (INIS)

Deffenbaugh, M.L.

1997-01-01

This document addresses the environmental permitting requirements for the transportation and interim storage of the Immobilized Low-Activity Waste (ILAW) produced during Phase 1 of the Hanford Site privatization effort. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage and disposal of Tank Waste Remediation Systems (TWRS) immobilized low-activity tank waste (ILAW) and (2) interim storage of TWRS immobilized HLW (IHLW) and other canistered high-level waste forms. Low-activity waste (LAW), low-level waste (LLW), and high-level waste (HLW) are defined by the TWRS, Hanford Site, Richland, Washington, Final Environmental Impact Statement (EIS) DOE/EIS-0189, August 1996 (TWRS, Final EIS). By definition, HLW requires permanent isolation in a deep geologic repository. Also by definition, LAW is ''the waste that remains after separating from high-level waste as much of the radioactivity as is practicable that when solidified may be disposed of as LLW in a near-surface facility according to the NRC regulations.'' It is planned to store/dispose of (ILAW) inside four empty vaults of the five that were originally constructed for the Group Program. Additional disposal facilities will be constructed to accommodate immobilized LLW packages produced after the Grout Vaults are filled. The specifications for performance of the low-activity vitrified waste form have been established with strong consideration of risk to the public. The specifications for glass waste form performance are being closely coordinated with analysis of risk. RL has pursued discussions with the NRC for a determination of the classification of the Hanford Site's low-activity tank waste fraction. There is no known RL action to change law with respect to onsite disposal of waste

Nuclear waste management. Quarterly progress report, April-June 1980

Energy Technology Data Exchange (ETDEWEB)

Platt, A.M.; Powell, J.A. (comps.)

1980-09-01

The status of the following programs is reported: high-level waste immobilization; alternative waste forms; Nuclear Waste Materials Characterization Center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; mobility of organic complexes of fission products in soils; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology; systems study on engineered barriers; criteria for defining waste isolation; spent fuel and fuel pool component integrity program; analysis of spent fuel policy implementation; asphalt emulsion sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; and development of backfill material.

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

International Nuclear Information System (INIS)

1991-05-01

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

Spent nuclear fuel and high level radioactive waste transportation. White paper

International Nuclear Information System (INIS)

1985-06-01

The High-Level Radioactive Waste Committee of the Western Interstate Energy Board has been involved in a year-long cooperative project with the US Department of Energy (DOE) to develop an information base on the transportation of spent nuclear fuel and high-level radioactive waste (HLW) so that western states can be constructive and informed participants in the repository program under the Nuclear Waste Policy Act (NWPA). The historical safety record of transportation of HLW and spent fuel is excellent; no release of these radioactive materials has ever occurred during transportation. Projected shipments under the NWPA will, however, greatly exceed current shipments in the US. For example, over the past five years, 119 metric tons of civilian spent fuel have been shipped in this country, while shipments to the first and second repository are each expected to peak at 3000 metric tons per year. The Committee believes that the successful development and operation of a national HLW/spent fuel transportation system can best be accomplished through an open process based on the common sense approach of taking all reasonable measures to minimize public risk and performing whatever actions are reasonably required to promote public acceptance. Therefore, the Committee recommends that the Department of Energy further the goals of the NWPA by developing a Comprehensive Transportation Plan which adopts a systematic, comprehensive, and integrated approach to resolving all spent fuel and HLW transportation issues in a timely manner. The suggested scope of such a plan is discussed in this White paper. Many of the suggested elements of such a plan are similar to those being developed by the Department of energy for inclusion in the Department's Transportation Institutional Plan

Disposition of surplus fissile materials via immobilization

International Nuclear Information System (INIS)

Gray, L.W.; Kan, T.; Sutcliffe, W.G.; McKibben, J.M.; Danker, W.

1995-01-01

In the Cold War aftermath, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, the USDOE has undertaken a multifaceted study to select options for storage and disposition of surplus plutonium (Pu). One disposition alternative being considered is immobilization. Immobilization is a process in which surplus Pu would be embedded in a suitable material to produce an appropriate form for ultimate disposal. To arrive at an appropriate form, we first reviewed published information on HLW immobilization technologies to identify forms to be prescreened. Surviving forms were screened using multi-attribute utility analysis to determine promising technologies for Pu immobilization. We further evaluated the most promising immobilization families to identify and seek solutions for chemical, chemical engineering, environmental, safety, and health problems; these problems remain to be solved before we can make technical decisions about the viability of using the forms for long-term disposition of Pu. All data, analyses, and reports are being provided to the DOE Office of Fissile Materials Disposition to support the Record of Decision that is anticipated in Summer of 1996

Immobilization of inorganic pyrophosphatase on nanodiamond particles retaining its high enzymatic activity.

Science.gov (United States)

Rodina, Elena V; Valueva, Anastasiya V; Yakovlev, Ruslan Yu; Vorobyeva, Nataliya N; Kulakova, Inna I; Lisichkin, Georgy V; Leonidov, Nikolay B

2015-12-21

Nanodiamond (ND) particles are popular platforms for the immobilization of molecular species. In the present research, enzyme Escherichia coli inorganic pyrophosphatase (PPase) was immobilized on detonation ND through covalent or noncovalent bonding and its enzymatic activity was characterized. Factors affecting adsorption of PPase such as ND size and surface chemistry were studied. The obtained material is a submicron size association of ND particles and protein molecules in approximately equal amounts. Both covalently and noncovalently immobilized PPase retains a significant enzymatic activity (up to 95% of its soluble form) as well as thermostability. The obtained hybrid material has a very high enzyme loading capacity (∼1 mg mg(-1)) and may be considered as a promising delivery system of biologically active proteinaceous substances, particularly in the treatment of diseases such as calcium pyrophosphate crystal deposition disease and related pathologies. They can also be used as recoverable heterogeneous catalysts in the traditional uses of PPase.

Should high-level nuclear waste be disposed of at geographically dispersed sites?

International Nuclear Information System (INIS)

Bassett, G.W. Jr.

1992-01-01

Consideration of the technical feasibility of Yucca Mountain in Nevada as the site for a high-level nuclear waste repository has led to an intense debate regarding the economic, social, and political impacts of the repository. Impediments to the siting process mean that the nuclear waste problem is being resolved by adhering to the status quo, in which nuclear waste is stored at scattered sites near major population centers. To assess the merits of alternative siting strategies--including both the permanent repository and the status quo- we consider the variables that would be included in a model designed to select (1) the optimal number of disposal facilities, (2) the types of facilities (e.g., permanent repository or monitored retrievable facility), and (3) the geographic location of storage sites. The objective function in the model is an all-inclusive measure of social cost. The intent of the exercise is not to demonstrate the superiority of any single disposal strategy; uncertainties preclude a conclusive proof of optimality for any of the disposal options. Instead, we want to assess the sensitivity of a variety of proposed solutions to variations in the physical, economic, political, and social variables that influence a siting strategy

Hierarchically Nanoporous Bioactive Glasses for High Efficiency Immobilization of Enzymes

DEFF Research Database (Denmark)

He, W.; Min, D.D.; Zhang, X.D.

2014-01-01

Bioactive glasses with hierarchical nanoporosity and structures have been heavily involved in immobilization of enzymes. Because of meticulous design and ingenious hierarchical nanostructuration of porosities from yeast cell biotemplates, hierarchically nanostructured porous bioactive glasses can...... and products of catalytic reactions can freely diffuse through open mesopores (2–40 nm). The formation mechanism of hierarchically structured porous bioactive glasses, the immobilization mechanism of enzyme and the catalysis mechanism of immobilized enzyme are then discussed. The novel nanostructure...

Co-Immobilization of Proteins and DNA Origami Nanoplates to Produce High-Contrast Biomolecular Nanoarrays.

Science.gov (United States)

Hager, Roland; Burns, Jonathan R; Grydlik, Martyna J; Halilovic, Alma; Haselgrübler, Thomas; Schäffler, Friedrich; Howorka, Stefan

2016-06-01

The biofunctionalization of nanopatterned surfaces with DNA origami nanostructures is an important topic in nanobiotechnology. An unexplored challenge is, however, to co-immobilize proteins with DNA origami at pre-determined substrate sites in high contrast relative to the nontarget areas. The immobilization should, in addition, preferably be achieved on a transparent substrate to allow ultrasensitive optical detection. If successful, specific co-binding would be a step towards stoichiometrically defined arrays with few to individual protein molecules per site. Here, we successfully immobilize with high specificity positively charged avidin proteins and negatively charged DNA origami nanoplates on 100 nm-wide carbon nanoislands while suppressing undesired adsorption to surrounding nontarget areas. The arrays on glass slides achieve unprecedented selectivity factors of up to 4000 and allow ultrasensitive fluorescence read-out. The co-immobilization onto the nanoislands leads to layered biomolecular architectures, which are functional because bound DNA origami influences the number of capturing sites on the nanopatches for other proteins. The novel hybrid DNA origami-protein nanoarrays allow the fabrication of versatile research platforms for applications in biosensing, biophysics, and cell biology, and, in addition, represent an important step towards single-molecule protein arrays. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

High-Level Waste Melter Study Report

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-13

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

The systems approach to spent fuel and high level nuclear waste management

International Nuclear Information System (INIS)

Norman, N.A.

1987-01-01

Based on the prior successful experience in the application of systems engineering techniques to large, complex, multi-participant programs the U.S. DOE has elected to apply the systems engineering process to the U.S. High Level Nuclear Waste Management Program (HLNWMP). Features of the systems engineering process as it applies to the U.S. HLNWMP and the importance of the component interdependencies are descirbed. Four principle components of the U.S. HLW Managment System are also given. They are the reactors or defense sources, transportation system, interium storoge modes, and the final repository. (Huang)

Systems approach to nuclear waste glass development

International Nuclear Information System (INIS)

Jantzen, C.M.

1986-01-01

Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan

Enzyme Technology of Peroxidases: Immobilization, Chemical and Genetic Modification

Science.gov (United States)

Longoria, Adriana; Tinoco, Raunel; Torres, Eduardo

An overview of enzyme technology applied to peroxidases is made. Immobilization on organic, inorganic, and hybrid supports; chemical modification of amino acids and heme group; and genetic modification by site-directed and random mutagenesis are included. Different strategies that were carried out to improve peroxidase performance in terms of stability, selectivity, and catalytic activity are analyzed. Immobilization of peroxidases on inorganic and organic materials enhances the tolerance of peroxidases toward the conditions normally found in many industrial processes, such as the presence of an organic solvent and high temperature. In addition, it is shown that immobilization helps to increase the Total Turnover Number at levels high enough to justify the use of a peroxidase-based biocatalyst in a synthesis process. Chemical modification of peroxidases produces modified enzymes with higher thermostability and wider substrate variability. Finally, through mutagenesis approaches, it is possible to produce modified peroxidases capable of oxidizing nonnatural substrates with high catalytic activity and affinity.

High-level nuclear waste borosilicate glass: A compendium of characteristics

International Nuclear Information System (INIS)

Cunnane, J.C.; Bates, J.K.; Ebert, W.L.; Feng, X.; Mazer, J.J.; Wronkiewicz, D.J.; Sproull, J.; Bourcier, W.L.; McGrail, B.P.

1992-01-01

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized

Research activity of institute of physical chemistry of Russian Academy of sciences in the field of nuclear fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Pikaev, A.K. [Institute of Physical Chemistry of Russian Academy of Sciences, Moscow (Russian Federation)

2000-07-01

The report is a brief review of the most important directions in research activity of the Institute of Physical Chemistry of RAS (Moscow) in the field of nuclear fuel cycle. The main attention is paid to researches and developments on liquid radioactive waste management including the removal of wastes to deep geological formations and the immobilization of the wastes. In particular, the data from the study on the properties of new, basaltic-like matrices for the immobilization are presented. The results of research on gas evolution from the systems modeling liquid high-level radioactive wastes are considered. The separation of some radionuclides from irradiated nuclear and the production of radiation sources by various methods are discussed. (author)

Basic approach to the disposal of low level radioactive waste generated from nuclear reactors containing comparatively high radioactivity

International Nuclear Information System (INIS)

Moriyama, Yoshinori

1998-01-01

Low level radioactive wastes (LLW) generated from nuclear reactors are classified into three categories: LLW containing comparatively high radioactivity; low level radioactive waste; very low level radioactive waste. Spent control rods, part of ion exchange resin and parts of core internals are examples of LLW containing comparatively high radioactivity. The Advisory Committee of Atomic Energy Commission published the report 'Basic Approach to the Disposal of LLW from Nuclear Reactors Containing Comparatively High Radioactivity' in October 1998. The main points of the proposed concept of disposal are as follows: dispose of underground deep enough not be disturb common land use (e.g. 50 to 100 m deep); dispose of underground where radionuclides migrate very slowly; dispose of with artificial engineered barrier which has the same function as the concrete pit; control human activities such as land use of disposal site for a few hundreds years. (author)

Immobilization of α-amylase and amyloglucosidase onto ion-exchange resin beads and hydrolysis of natural starch at high concentration.

Science.gov (United States)

Gupta, Kapish; Jana, Asim Kumar; Kumar, Sandeep; Maiti, Mithu

2013-11-01

α-Amylase was immobilized on Dowex MAC-3 with 88 % yield and amyloglucosidase on Amberlite IRA-400 ion-exchange resin beads with 54 % yield by adsorption process. Immobilized enzymes were characterized to measure the kinetic parameters and optimal operational parameters. Optimum substrate concentration and temperature were higher for immobilized enzymes. The thermal stability of the enzymes enhanced after the immobilization. Immobilized enzymes were used in the hydrolysis of the natural starch at high concentration (35 % w/v). The time required for liquefaction of starch to 10 dextrose equivalent (DE) and saccharification of liquefied starch to 96 DE increased. Immobilized enzymes showed the potential for use in starch hydrolysis as done in industry.

Supported liquid inorganic membranes for nuclear waste separation

Science.gov (United States)

Bhave, Ramesh R; DeBusk, Melanie M; DelCul, Guillermo D; Delmau, Laetitia H; Narula, Chaitanya K

2015-04-07

A system and method for the extraction of americium from radioactive waste solutions. The method includes the transfer of highly oxidized americium from an acidic aqueous feed solution through an immobilized liquid membrane to an organic receiving solvent, for example tributyl phosphate. The immobilized liquid membrane includes porous support and separating layers loaded with tributyl phosphate. The extracted solution is subsequently stripped of americium and recycled at the immobilized liquid membrane as neat tributyl phosphate for the continuous extraction of americium. The sequestered americium can be used as a nuclear fuel, a nuclear fuel component or a radiation source, and the remaining constituent elements in the aqueous feed solution can be stored in glassified waste forms substantially free of americium.

Impacts of biochar and oyster shells waste on the immobilization of arsenic in highly contaminated soils.

Science.gov (United States)

Chen, Yongshan; Xu, Jinghua; Lv, Zhengyong; Xie, Ruijia; Huang, Liumei; Jiang, Jinping

2018-07-01

Soil contamination is a serious problem with deleterious impacts on global sustainability. Readily available, economic, and highly effective technologies are therefore urgently needed for the rehabilitation of contaminated sites. In this study, two readily available materials prepared from bio-wastes, namely biochar and oyster shell waste, were evaluated as soil amendments to immobilize arsenic in a highly As-contaminated soil (up to 15,000 mgAs/kg). Both biochar and oyster shell waste can effectively reduce arsenic leachability in acid soils. After application of the amendments (2-4% addition, w/w), the exchangeable arsenic fraction decreased from 105.8 to 54.0 mg/kg. The application of 2%biochar +2% oyster shell waste most effectively reduced As levels in the column leaching test by reducing the arsenic concentration in the porewater by 62.3% compared with the treatment without amendments. Biochar and oyster shell waste also reduced soluble As(III) from 374.9 ± 18.8 μg/L to 185.9 ± 16.8 μg/L and As(V) from 119.8 ± 13.0 μg/L to 56.4 ± 2.6 μg/L at a pH value of 4-5. The treatment using 4% (w/w) amendments did not result in sufficient As immobilization in highly contaminated soils; high soluble arsenic concentrations (upto193.0 μg/L)were found in the soil leachate, particularly in the form of As(III), indicating a significant potential to pollute shallow groundwater aquifers. This study provides valuable insights into the use of cost-effective and readily available materials for soil remediation and investigates the mechanisms underlying arsenic immobilization in acidic soils. Copyright © 2018 Elsevier Ltd. All rights reserved.

High level waste management in Asia: R and D perspectives

International Nuclear Information System (INIS)

Deokattey, Sangeeta; Bhanumurthy, K.

2010-01-01

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

Conditioning and storage of low level radioactive waste in FR Yugoslavia

International Nuclear Information System (INIS)

Plecas, I.; Pavlovic, R.; Pavlovic, S.

2000-01-01

FR Yugoslavia is a country without any nuclear power plant on its territory. In the last forty years in the country, as a result of the two research reactors operation and also from radionuclides applications in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity are generated. As a temporary solution, these radioactive waste materials are stored in the two interim storage facility. Since one of the storage facilities is completely full with radioactive wastes, packed in metal drums and plastic barrels, and the second one has an effective space for the next few years, attempts are made in the 'Vinca' Institute of Nuclear Sciences in developing the the immobilization process, for low and intermediate level radioactive wastes and their safe disposal. As an immobilization process, cementation process is investigated. Developed immobilization process has, as a final goal, production of solidified waste-matrix mixture form, that is easy for handling and satisfies requirements for interim storage and final disposal. Radioactive wastes immobilized in inactive matrices are to be placed into concrete containers for further manipulation and disposal

High-level radioactive waste management. A means to social consensus

International Nuclear Information System (INIS)

Pierce, B.; Hill, D.

1984-01-01

The problem of safely disposing of high-level radioactive waste is not new, but it is becoming more pressing as the temporary storage facilities of public utilities run out. The technical questions of how best to immobilize these wastes for many centuries have been studied for years and many feel that these problems are solved, or nearly so. In the USA many states have set up roadblocks to the federal waste management programme, however, and it is clear that social consensus must be reached for any waste disposal programme to be successful. The Nuclear Waste Policy Act of 1982 provides a long needed framework for reaching this consensus, giving the states unprecedented access to federal decision making. The rights of the states in a process of co-operation and consultation are clearly defined by the Act, but the means by which the states exercise these rights are left entirely to them. We examine the structures, methods and goals open to the states, and recommend a rationale for the state decision process defining the roles of the governor and legislature. (author)

High-level radioactive wste management: a means to social consensus

International Nuclear Information System (INIS)

Pierce, B.; Hill, D.; Haefele, E.T.

1983-01-01

The problem of safely disposing of high-level radioactive waste is not new, but it is becoming more pressing as the temporary storage facilities of public utilities run out. The technical questions of how best to immobilize these wastes for many centuries have been studied for years and many feel that these problems are solved, or nearly so. Many states have set up roadblocks to the federal waste management program, however, and it is clear that social consensus must be reached for any waste disposal program to be successful. The Nuclear Waste Policy Act of 1982 provides a long-needed framework for reaching this consensus, giving the states unprecedented access to federal decision-making. The rights of the states in a process of cooperation and consultation are clearly defined by the Act, but the means by which the states exercise those rights are left entirely to them. We examine the structures, methods, and goals open to the states, and recommend a rationale for the state decision process defining the roles of the governor and legislature

Mechanical stability of bentonite buffer system for high level nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Lempinen, A. [Helsinki Univ. of Technology, Espoo (Finland). Lab. of Theoretical and Applied Mechanics

1998-05-01

According to present plans, high level nuclear waste in Finland is going to be disposed of in bedrock at a depth of several hundred metres. The spent fuel containers will be placed in boreholes drilled in the floors of deposition tunnels with engineered clay buffer, which is made of bentonite blocks. The tunnels will be filled with a mixture of bentonite and crushed rock. For stability calculations a thermomechanical model for compressed bentonite is needed. In the study a thermomechanically consistent model for reversible processes for swelling clays is presented. Preliminary calculations were performed and they show that uncertainty in material parameter values causes significantly different results. Therefore, measurements that are consistent with the model are needed 12 refs.

Mechanical stability of bentonite buffer system for high level nuclear waste

International Nuclear Information System (INIS)

Lempinen, A.

1998-05-01

According to present plans, high level nuclear waste in Finland is going to be disposed of in bedrock at a depth of several hundred metres. The spent fuel containers will be placed in boreholes drilled in the floors of deposition tunnels with engineered clay buffer, which is made of bentonite blocks. The tunnels will be filled with a mixture of bentonite and crushed rock. For stability calculations a thermomechanical model for compressed bentonite is needed. In the study a thermomechanically consistent model for reversible processes for swelling clays is presented. Preliminary calculations were performed and they show that uncertainty in material parameter values causes significantly different results. Therefore, measurements that are consistent with the model are needed

Site investigations for final disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Aeikaes, T.; Laine, T.

1982-12-01

Research concerning disposal of high-level nuclear waste of the Industrial Power Company Ltd has focused on deep underground disposal in Finnish precambrian bedrock. The present target is to have a repository for high-level waste in operation by 2020. Selection of the repository site is based on site investigations. In addition to geosciences, selection of appropriate site includes many branches of studies; engineering, safety analysis, ecology, transport, demography etc. The investigations required for site selection for high-level waste have been arranged in a sequence of four phases. The aim of the phases is that investigations become more and more detailed as the selection process continues. Phase I of the investigations is the characterization of potential areas. This comprises establishment of criteria for site selection and identification of areas that meet selection criteria. Objective of these studies is to determine areas for phase II field investigations. The studies are largely made by reviewing existing data and remote-sensing techniques. Phase II field investigations will be undertaken between 1986-1992. The number of potential candidates for repository site is reduced to few preferred areas by preceeding generic study. The site selection process culminates in phase III in site confirmation studies carried out at 2...3 most suitable sites during 1992-2010. This is then followed by phase IV, which comprises very detailed investigations at the selected site. An alternative for these investigations is to undertake them by using pilot shaft and drifts. Active development is taking place in all phases concerning investigation methods, criteria, parameters, data processing and modelling. The applicability of the various investigation methods and techniques is tested in a deep borehole in phase I. The co-operation with countries with similar geological conditions makes it possible to compare results obtained by different techniques

PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION

Energy Technology Data Exchange (ETDEWEB)

J.A. Ziegler

2000-11-20

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

Legal and judicial perspectives on the disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Muntzing, L.M.

1985-01-01

Nuclear Waste Policy Act of 1982 creates a maze of legal requirements that are complex and subject to differing interpretations. The intervention of the courts should be anticipated as varying interests dispute the correct path through the maze to high-level waste disposal. A significant number of legal issues and unsettled questions exist that will need to be resolved. The challenge will be to minimize the delays that legal conflicts can produce. This can be helped by resolving uncertainties and conflicts early before they are on the critical path or negotiating resolutions, normally a difficult and costly endeavor

Improved polyphase ceramic form for high-level defense nuclear waste

International Nuclear Information System (INIS)

Harker, A.B.; Morgan, P.E.D.; Clarke, D.R.; Flintoff, J.J.; Shaw, T.M.

1983-01-01

An improved ceramic nuclear waste form and fabrication process have been developed using simulated Savannah River Plant defense high-level waste compositions. The waste form provides flexibility with respect to processing conditions while exhibiting superior resistance to ground water leaching than other currently proposed forms. The ceramic, consolidated by hot-isostatic pressing at 1040 0 C and 10,000 psi, is composed of six major phases, nepheline, zirconolite, a murataite-type cubic phase, magnetite-type spinel, a magnetoplumbite solid solution, and perovskite. The waste form provides multiple crystal lattice sites for the waste elements, minimizes amorphous intergranular material, and can accommodate waste loadings in excess of 60 wt %. The fabrication of the ceramic can be accomplished with existing manufacturing technology and eliminates the effects of radionuclide volatilization and off-gas induced corrosion experienced with the molten processes for vitreous form production

Thermo-mechanical analysis of high level nuclear wastes in granite

International Nuclear Information System (INIS)

Millard, A.; Guri, G.; Raimbault, M.

1991-01-01

In order to appraise the safety of a storage of high level nuclear wastes in rock masses, it is necessary to assess, among other features, the thermo-mechanical behaviour of the host rock for long periods (thousands of years). In France, four different media are considered as potential host rocks: granite, shale, salt, clay. The present paper is devoted to some analysis of a generic storage configuration in granite. The case of a rock mass without any major fault has been considered. The granite is modelled by means of an elastic fracturing model (no tension type). The results obtained show that some fissures, induced by the heat generation, develop mainly above the repository. The opening of the fissures, within the frame of the adopted hypothesis, have not a strong influence on the rock mass, as a geological barrier for the radionuclides. (author)

Process for solidifying high-level nuclear waste

Science.gov (United States)

Ross, Wayne A.

1978-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

No public policy issue has been as difficult as high-level nuclear waste. Debates continue regarding Yucca Mountain as a disposal site, and - more generally - the appropriateness of geologic disposal and the need to act quickly. Previous research has focused on possible social, political, and economic consequences of a facility in Nevada. Impacts have been predicted to be potentially large and to emanate mainly from stigmatization of the region due to increased perceptions of risk. Analogous impacts from leaving waste at power plants have been either ignored or assumed to be negligible. This paper presents survey results on attitudes of residents in three countries where nuclear waste is currently stored. Topics include perceived risk, knowledge of nuclear waste and radiation, and impacts on jobs, tourism, and housing values from leaving waste on site. Results are similar to what has been reported for Nevada; the public is concerned about possible adverse effects from on-site storage of waste. 24 refs., 7 figs., 5 tabs

Can Sisyphus succeed? Getting U.S. high-level nuclear waste into a geological repository.

Science.gov (United States)

North, D Warner

2013-01-01

The U.S. government has the obligation of managing the high-level radioactive waste from its defense activities and also, under existing law, from civilian nuclear power generation. This obligation is not being met. The January 2012 Final Report from the Blue Ribbon Commission on America's Nuclear Future provides commendable guidance but little that is new. The author, who served on the federal Nuclear Waste Technical Review Board from 1989 to 1994 and subsequently on the Board on Radioactive Waste Management of the National Research Council from 1994 to 1999, provides a perspective both on the Commission's recommendations and a potential path toward progress in meeting the federal obligation. By analogy to Sisyphus of Greek mythology, our nation needs to find a way to roll the rock to the top of the hill and have it stay there, rather than continuing to roll back down again. © 2012 Society for Risk Analysis.

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

Science.gov (United States)

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

1996-06-01

No public policy issue has been as difficult as high-level nuclear waste. Debates continue regarding Yucca Mountain as a disposal site, and-more generally-the appropriateness of geologic disposal and the need to act quickly. Previous research has focused on possible social, political, and economic consequences of a facility in Nevada. Impacts have been predicted to be potentially large and to emanate mainly from stigmatization of the region due to increased perceptions of risk. Analogous impacts from leaving waste at power plants have been either ignored or assumed to be negligible. This paper presents survey results on attitudes of residents in three counties where nuclear waste is currently stored. Topics include perceived risk, knowledge of nuclear waste and radiation, and impacts on jobs, tourism, and housing values from leaving waste on site. Results are similar to what has been reported for Nevada; the public is concerned about possible adverse effects from on-site storage of waste.

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

International Nuclear Information System (INIS)

Anon.

1978-01-01

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

Retrievability of high level waste and spent nuclear fuel. Proceedings of an international seminar

International Nuclear Information System (INIS)

2000-12-01

The possibility of retrieving spent nuclear fuel or reprocessing high-level radioactive wastes placed in geological repositories is an issue that has attracted increased attention during the past few years, not only among technical experts but also among politicians at different levels, environmental organisations and other interested representatives of the public. This publication contains the presented invited papers, an edited record of the discussions and some concluding remarks. The seminar addressed a wide range of aspects of retrievability including technical options; public acceptance; ethical aspects; long term monitoring and cost considerations; safety and regulatory aspects. Each of the presented papers was indexed separately

Retrievability of high level waste and spent nuclear fuel. Proceedings of an international seminar

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-12-01

The possibility of retrieving spent nuclear fuel or reprocessing high-level radioactive wastes placed in geological repositories is an issue that has attracted increased attention during the past few years, not only among technical experts but also among politicians at different levels, environmental organisations and other interested representatives of the public. This publication contains the presented invited papers, an edited record of the discussions and some concluding remarks. The seminar addressed a wide range of aspects of retrievability including technical options; public acceptance; ethical aspects; long term monitoring and cost considerations; safety and regulatory aspects. Each of the presented papers was indexed separately.

An overview of technologies for immobilization of enzymes and surface analysis techniques for immobilized enzymes

Science.gov (United States)

Mohamad, Nur Royhaila; Marzuki, Nur Haziqah Che; Buang, Nor Aziah; Huyop, Fahrul; Wahab, Roswanira Abdul

2015-01-01

The current demands of sustainable green methodologies have increased the use of enzymatic technology in industrial processes. Employment of enzyme as biocatalysts offers the benefits of mild reaction conditions, biodegradability and catalytic efficiency. The harsh conditions of industrial processes, however, increase propensity of enzyme destabilization, shortening their industrial lifespan. Consequently, the technology of enzyme immobilization provides an effective means to circumvent these concerns by enhancing enzyme catalytic properties and also simplify downstream processing and improve operational stability. There are several techniques used to immobilize the enzymes onto supports which range from reversible physical adsorption and ionic linkages, to the irreversible stable covalent bonds. Such techniques produce immobilized enzymes of varying stability due to changes in the surface microenvironment and degree of multipoint attachment. Hence, it is mandatory to obtain information about the structure of the enzyme protein following interaction with the support surface as well as interactions of the enzymes with other proteins. Characterization technologies at the nanoscale level to study enzymes immobilized on surfaces are crucial to obtain valuable qualitative and quantitative information, including morphological visualization of the immobilized enzymes. These technologies are pertinent to assess efficacy of an immobilization technique and development of future enzyme immobilization strategies. PMID:26019635

Studies on the immobilization of simulated HLW in NaTi2(PO4)3 (NTP) matrix

International Nuclear Information System (INIS)

Raja Madhavan, R.; Govindan Kutty, K.V.; Gandhi, A.S.

2015-01-01

Immobilization of high level nuclear waste (HLW) is a big challenge faced by the nuclear industry today. The HLW has to be contained and isolated from the biosphere for geological timescales. NZP family of compounds is very versatile monophasic hosts for HLW immobilization. Their crystal structure can accommodate nearly all the cations known to be present in HLW due to its open structure with voids of different size. In the present study a systematic investigation on NaTi 2 (PO 4 ) 3 belonging to the NZP family; as a potential host for HLW immobilization was carried out. A simulated HLW expected from Fast Breeder Test Reactor, India (FBTR) (150Gwd/T burnup, 1 year cooling) was used. Simulated NTP waste forms with 5, 10, 15 wt. % waste loading were prepared by employing a wet chemical method and characterized. Single phase simulated NTP waste forms with up to 5 wt.% waste loading could be prepared for samples sintered in air and above 5 wt.% waste loading, monazite phase is observed as a minor secondary phase. It was found that when sintering is done in Ar/10%H 2 , NTP matrix accepts up to 10 wt.% waste loading without formation of any second phase. From the SEM studies, it was observed that samples sintered in air as well as Ar/10%H 2 palladium segregated as a metal phase and uniformly distributed throughout the waste matrix. The elemental mapping revealed retention of some of the fission products like Ru, Mo, Cs that are volatile during sintering above 1173 K and are homogenously distributed in the matrix. (author)

Corrosion of steel drums containing immobilized ion exchange-resins and incineration ashes

International Nuclear Information System (INIS)

Marotta, F.; Schulz Rodriguez, F.M.; Farina, Silvia B.; Duffo, Gustavo S.

2009-01-01

The Argentine Atomic Energy Commission (CNEA) is responsible for developing the management nuclear waste disposal programme. This programme contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The proposed model is a near-surface monolithic repository similar to those in operation in El Cabril, Spain. The design of this type of repository is based on the use of multiple, independent and redundant barriers. The intermediate radioactive waste consists mostly in spent ionic exchange resins and filters from the nuclear power plants, research reactors and radioisotopes production facilities. The spent resins, as well as the incineration ashes, have to be immobilized before being stored to improve leach resistance of waste matrix and to maintain mechanical stability for safety requirements. Generally, cementation processes have been used as immobilization techniques for economical reasons as well as for being a simple operation. The immobilized resins and incineration ashes are thus contained in steel drums that, in turn, can undergo corrosion depending on the ionic content of the matrix. This work is a part of a systematic study of the corrosion susceptibility of steel drums in contact with immobilized cemented exchange-resins with different types and contents of aggressive species and incineration ashes. To this purpose, a special type of specimen was manufactured to simulate the cemented waste in the drum. The evolution of the corrosion potential and the corrosion current density of the steel, as well as the electrical resistivity of the matrix are being monitored along time. The aggressive species studied were chloride ions (the main ionic species present in nature) and sulphate ions (produced during the radiolysis process of the cationic exchange-resins after cementation). Preliminary results show the strong effect of chloride on the corrosion susceptibility of the steel. Monitoring will continue for

Effects of immobilization and whole-body vibration on rat serum Type I collagen turnover.

Science.gov (United States)

Dönmez, Gürhan; Doral, Mahmut Nedim; Suljevic, Şenay; Sargon, Mustafa Fevzi; Bilgili, Hasan; Demirel, Haydar Ali

2016-08-01

The aim of this study was to investigate the effects of short-term, high-magnitude whole-body vibration (WBV) on serum type I collagen turnover in immobilized rats. Thirty Wistar albino rats were randomly divided into the following 5 groups: immobilization (IS), immobilization + remobilization (IR), immobilization + WBV (IV), control (C), and WBV control (CV). Immobilization was achieved by casting from the crista iliaca anterior superior to the lower part of the foot for 2 weeks. The applied WBV protocol involved a frequency of 45 Hz and amplitude of 3 mm for 7 days starting a day after the end of the immobilization period. Serum type I collagen turnover markers were measured by using ELISA kits. Serum NH2-terminal propeptide of type I collagen (PINP) levels were significantly lower in the immobilization groups (p immobilization groups. Similarly, serum COOH-terminal telopeptide of type I collagen (CTX) levels were higher in the WBV controls than their own controls (p Immobilization led to deterioration of tendon tissue, as observed by histopathological analysis with a transmission electron microscope. Although 1 week of WBV had a positive effect on type I collagen turnover in controls, it is not an efficient method for repairing tissue damage in the early stage following immobilization. Copyright © 2016 Turkish Association of Orthopaedics and Traumatology. Production and hosting by Elsevier B.V. All rights reserved.

Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

International Nuclear Information System (INIS)

Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

2013-01-01

A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl 11 O 18 and Ce 2 SiO 5 . The leaching rate of cerium over a period of 28 days was 10 −5 –10 −6 g/(m 2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products

Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

Science.gov (United States)

Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

2013-11-01

A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl11O18 and Ce2SiO5. The leaching rate of cerium over a period of 28 days was 10-5-10-6 g/(m2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products.

Immobilization of preconditioned spent fuel from nuclear research reactors in a ceramic matrix

International Nuclear Information System (INIS)

Russo, Diego O.; Rodriguez, Diego S.; Heredia, Arturo D.; Sanfilippo, Miguel; Sterba, Mario E.; Mateos, Patricia

2002-01-01

The fuel elements from nuclear research reactors consist in a laminated sandwich of aluminum with a core of some uranium compound. To process this material its necessary to previously eliminate the aluminum covering the fuel, before the conditioning of the rest of the fuel in a stable matrix, in order to obtain an acceptable waste form for a subsequent disposition in a geological repository. Normally, mechanical and chemical methods are proposed for that purpose. One of the most developed techniques for immobilization of the radioactive elements above mentioned, is the vitrification. In this work we propose a method named CERUS (in Spanish Ceramizacion de Elementos Radiactivos con Uranio Sinterizado - Ceramization of radioactive elements with sintered uranium). This is a sinterization of the pre-treated fuel elements mixed with natural uranium oxide. The properties of the blocks obtained are adequate for final disposal in a deep geological reservoir. (author)

High-level waste melter alternatives assessment report

Energy Technology Data Exchange (ETDEWEB)

Calmus, R.B.

1995-02-01

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

High-level waste melter alternatives assessment report

International Nuclear Information System (INIS)

Calmus, R.B.

1995-02-01

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

Amperometric Determination of Glucose at Parts per Million Levels with Immobilized Glucose Oxidase.

Science.gov (United States)

Sittampalam, G.; Wilson, G. S.

1982-01-01

An experiment on the operation and utility of an amperometric immobilized enzyme electrode (or probe) is described, including advantages of the experiment, equipment, reagents, preparation of phosphate buffer, enzyme immobilization techniques, laboratory procedures, precautions, and discussion of experimental results. (SK)

Ecological risk assessment of deep geological disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Hart, D.R.; Lush, D.L.; Acton, D.W.

1993-01-01

Contaminant fate and transport models, radiological dosimetry models, chemical dose-response models and population dynamic models were used to estimate ecological risks to moose and brook trout populations arising from a proposed high-level nuclear waste repository. Risks from potential contaminant releases were compared with risks from physical habitat alteration in constructing a repository and service community, and with risks from increased hunting and fish pressure in the area. For a reference environment typical of a proposed location somewhere in the Canadian Shield, preliminary results suggest that the population consequences of contaminant release will be minor relative to those of habitat alteration and natural resource use

Defense High Level Waste Disposal Container System Description Document

International Nuclear Information System (INIS)

Pettit, N. E.

2001-01-01

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

Immobilization of low and intermediate level radioactive liquid wastes using some industrial by-product materials

International Nuclear Information System (INIS)

Sami, N.M.; EI-Dessouky, M.I.; Abou EI-Nour, F.H.; Abdel-Khalik, M.

2006-01-01

Immobilization of low and intermediate level.radioactive liquid wastes in different matrices: ordinary Portland cement and cement mixed with some industrial byproduct: by-pass kiln cement dust, blast furnace slag and ceramic sludge was studied. The effect of these industrial by-product materials on the compressive strength, water immersion, radiation effect and teachability were investigated. The obtained results showed that, these industrial by-product improve the cement pastes where they increase the compressive strength, decrease the leaching rate for radioactive cesium-137 and cobalt-60 ions through the solidified waste forms and increase resistance for y-radiation. It is found that, solidified waste forms of intermediate level liquid waste (ILLW) had high compressive strength values more than those obtained from low level liquid waste (LLLW). The compressive strength increased after immersion in different leachant for one and three months for samples with LLLW higher than those obtained for ILLW. The cumulative fractions released of cesium-137 and cobalt-60 of solidified waste forms of LLLW was lower than those obtained for ILLW

Synroc tailored waste forms for actinide immobilization

Energy Technology Data Exchange (ETDEWEB)

Gregg, Daniel J.; Vance, Eric R. [Australian Nuclear Science and Technology Organisation, Kirrawee (Australia). ANSTOsynroc, Inst. of Materials Engineering

2017-07-01

Since the end of the 1970s, Synroc at the Australian Nuclear Science and Technology Organisation (ANSTO) has evolved from a focus on titanate ceramics directed at PUREX waste to a platform waste treatment technology to fabricate tailored glass-ceramic and ceramic waste forms for different types of actinide, high- and intermediate level wastes. The particular emphasis for Synroc is on wastes which are problematic for glass matrices or existing vitrification process technologies. In particular, nuclear wastes containing actinides, notably plutonium, pose a unique set of requirements for a waste form, which Synroc ceramic and glass-ceramic waste forms can be tailored to meet. Key aspects to waste form design include maximising the waste loading, producing a chemically durable product, maintaining flexibility to accommodate waste variations, a proliferation resistance to prevent theft and diversion, and appropriate process technology to produce waste forms that meet requirements for actinide waste streams. Synroc waste forms incorporate the actinides within mineral phases, producing products which are much more durable in water than baseline borosilicate glasses. Further, Synroc waste forms can incorporate neutron absorbers and {sup 238}U which provide criticality control both during processing and whilst within the repository. Synroc waste forms offer proliferation resistance advantages over baseline borosilicate glasses as it is much more difficult to retrieve the actinide and they can reduce the radiation dose to workers compared to borosilicate glasses. Major research and development into Synroc at ANSTO over the past 40 years has included the development of waste forms for excess weapons plutonium immobilization in collaboration with the US and for impure plutonium residues in collaboration with the UK, as examples. With a waste loading of 40-50 wt.%, Synroc would also be considered a strong candidate as an engineered waste form for used nuclear fuel and highly

Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics

Energy Technology Data Exchange (ETDEWEB)

Vienna, John D.; Crum, Jarrod V.; Sevigny, Gary J.; Smith, G L.

2012-09-30

A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass ceramics waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan to conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1 • Feeding, melting, and pouring: TRL-1 • Glass ceramic formulation: TRL-1 • Canister cooling and crystallization: TRL-1 • Canister decontamination: TRL-4 Although the TRL’s are low for most of these CTE’s (TRL-1), the effort required to advance them to higher values. The activities required to advance the TRL’s are listed below: • Complete this TMP • Perform a preliminary engineering study • Characterize, estimate, and simulate waste to be treated • Laboratory scale glass ceramic testing • Melter and off-gas testing with simulants • Test the mixing, sampling, and analyses • Canister testing • Decontamination system testing • Issue a requirements document • Issue a risk management document • Complete preliminary design • Integrated pilot testing • Issue a waste compliance plan A preliminary schedule and budget were developed to complete these activities as summarized in the following table (assuming 2012 dollars). TRL Budget Year MSA FMP GCF CCC CD Overall $M 2012 1 1 1 1 4 1 0.3 2013 2 2 1 1 4 1 1.3 2014 2 3 1 1 4 1 1.8 2015 2 3 2 2 4 2 2.6 2016 2 3 2 2 4 2 4

Water level monitoring device in nuclear reactor

International Nuclear Information System (INIS)

Miura, Kiyohide; Otake, Tomohiro.

1988-01-01

Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

Immobilization of iodine in concrete

International Nuclear Information System (INIS)

Clark, W.E.; Thompson, C.T.

1977-01-01

A method for immobilizing fission product radioactive iodine recovered from irradiated nuclear fuel comprises combining material comprising water, Portland cement and about 3 to 20 wt percent iodine as Ba(IO 3 ) 2 to provide a fluid mixture and allowing the fluid mixture to harden, said Ba(IO 3 ) 2 comprising said radioactive iodine. An article for solid waste disposal comprises concrete prepared by this method. 10 claims, 2 figures

Short-Term Limb Immobilization Affects Cognitive Motor Processes

Science.gov (United States)

Toussaint, Lucette; Meugnot, Aurore

2013-01-01

We examined the effects of a brief period of limb immobilization on the cognitive level of action control. A splint placed on the participants' left hand was used as a means of immobilization. We used a hand mental rotation task to investigate the immobilization-induced effects on motor imagery performance (Experiments 1 and 2) and a number mental…

Cervical spine fractures in elderly patients with hip fracture after low-level fall: an opportunity to refine prehospital spinal immobilization guidelines?

Science.gov (United States)

Boland, Lori L; Satterlee, Paul A; Jansen, Paul R

2014-02-01

Conventional prehospital spine-assessment approaches based on low index of suspicion and mechanism of injury (MOI) result in the liberal application of spinal immobilization in trauma patients. A painful distracting injury (DI), such as a suspected hip fracture, historically has been a sufficient condition for immobilization, even in an elderly patient who suffers a simple fall from standing and exhibits no other risk factors for spinal injury. Because the elderly are at increased risk of hip fracture from low-level falls, and are also particularly susceptible to the discomfort and morbidity associated with immobilization, the prevalence of cervical spine (c-spine) fracture in this patient population was examined. Hospital billing records were used to identify all cases of traumatic femur fracture in Minnesota (USA) in 2010-2011. Concurrent diagnosis and external cause codes were used to estimate the prevalence of c-spine fracture by age and MOI. Among 1,394 patients with femur fracture, 23 (1.7%) had a c-spine fracture. When the MOI was a fall from standing or sitting height and the patient age was ≥ 65, the prevalence dropped to 0.4% (2/565). The prevalence was similar when the definition of hip fracture additionally included pelvis fractures (0.5%; 11/2,441). Eight of the 11 patients with c-spine fracture had diagnosis codes indicative of criteria other than the DI that likely would have resulted in immobilization (eg, head injury and compromised mental status). C-spine fracture is extremely rare in elderly patients who sustain hip fracture as a result of a low-level fall, and appears to be accompanied frequently by other known predictors of spinal injury besides DI. More research is needed to determine whether conservative use of spinal immobilization may be warranted in elderly patients with hip fracture after low-level falls when the only criteria for immobilization is the distracting hip injury.

HLW immobilization in glass: industrial operation and product quality

International Nuclear Information System (INIS)

Jacquet-Francillon, N.; Leroy, P.; Runge, S.

1992-01-01

This extended summary discusses the immobilization of high level wastes from the viewpoint of the quality of the final product, i.e. the HLW glass. The R and D studies comprise 3 steps: glass formulation, glass characterization and long term behaviour studies

Immobilization of {sup 99}Tc (Re) using Iron-Phosphate Glasses

Energy Technology Data Exchange (ETDEWEB)

Heo, Jong; Xu, Kai; Um, Woo Yong; Hrma, Pavel [Pohang Univ. of Science and Technology, Pohang (Korea, Republic of)

2012-05-15

Technetium-99 ({sup 99}Tc) is a fission product artificially generated during the irradiation of {sup 235}U for commercial power production or {sup 239}Pu for nuclear weapons. Under oxidizing conditions, the dominant species of Tc, the pertechnetate anion (TcO{sub 4} {sup -}), is highly soluble in ground water and thus easily transports through the geologic systems. In addition, because of its high fission yield ({approx}6 %) and long half-life (2.1x10{sup 5} yr), immobilization of {sup 99}Tc has been investigated for decades. Several waste forms such as metallic alloys, sintered titanate ceramics and chemically bonded phosphate ceramics have been proposed to encapsulate {sup 99}Tc. They have not yet been realized in the industrial-scale, mostly either due to the high volatilization of {sup 99}Tc during high temperature process (>1300 .deg. C), or the low {sup 99}Tc loading. Iron-phosphate (FeP) glasses have been developed as alternative waste forms because of their chemical durability equivalent to borosilicate glasses. Additionally, vitrification of radioactive waste by FeP glasses can be done at a relatively low temperature ({approx}1000 .deg. C) and the low-temperature process can reduce the volatilization of {sup 99}Tc significantly. Thus, this work reports the immobilization of {sup 99}Tc by FeP glasses using rhenium (Re) as a surrogate. We also examine the chemical durability of Re-containing FeP glasses using product consistency test (PCT). Experimental results reveal that FeP glass can become a promising candidate for immobilizing {sup 99}Tc

Surface cell immobilization within perfluoroalkoxy microchannels

Energy Technology Data Exchange (ETDEWEB)

Stojkovič, Gorazd; Krivec, Matic [Faculty of Chemistry and Chemical Technology, University of Ljubljana, Aškerčeva 5, SI-1000 Ljubljana (Slovenia); Vesel, Alenka [Jožef Stefan Institute, Jamova cesta 39, 1000 Ljubljana (Slovenia); Marinšek, Marjan [Faculty of Chemistry and Chemical Technology, University of Ljubljana, Aškerčeva 5, SI-1000 Ljubljana (Slovenia); Žnidaršič-Plazl, Polona, E-mail: polona.znidarsic@fkkt.uni-lj.si [Faculty of Chemistry and Chemical Technology, University of Ljubljana, Aškerčeva 5, SI-1000 Ljubljana (Slovenia)

2014-11-30

Graphical abstract: - Highlights: • A very efficient approach for immobilization of cells into microreactors is presented. • It is applicable to various materials, including PFA and cyclic olefin (co)polymers. • It was used to immobilize different prokaryotic and eukaryotic microbes. • Cells were immobilized on the surface in high density and showed good stability. • Mechanisms of APTES interactions with target materials are proposed. - Abstract: Perfluoroalkoxy (PFA) is one of the most promising materials for the fabrication of cheap, solvent resistant and reusable microfluidic chips, which have been recently recognized as effective tools for biocatalytic process development. The application of biocatalysts significantly depends on efficient immobilization of enzymes or cells within the reactor enabling long-term biocatalyst use. Functionalization of PFA microchannels by 3-aminopropyltriethoxysilane (ATPES) and glutaraldehyde was used for rapid preparation of microbioreactors with surface-immobilized cells. X-ray photoelectron spectroscopy and scanning electron microscopy were used to accurately monitor individual treatment steps and to select conditions for cell immobilization. The optimized protocol for Saccharomyces cerevisiae immobilization on PFA microchannel walls comprised ethanol surface pretreatment, 4 h contacting with 10% APTES aqueous solution, 10 min treatment with 1% glutaraldehyde and 20 min contacting with cells in deionized water. The same protocol enabled also immobilization of Escherichia coli, Pseudomonas putida and Bacillus subtilis cells on PFA surface in high densities. Furthermore, the developed procedure has been proved to be very efficient also for surface immobilization of tested cells on other materials that are used for microreactor fabrication, including glass, polystyrene, poly (methyl methacrylate), polycarbonate, and two olefin-based polymers, namely Zeonor{sup ®} and Topas{sup ®}.

DWPF upgrade, immobilization Programmatic Environmental Impact Statement input. Revision 1

International Nuclear Information System (INIS)

Sullivan, I.K.; Bignell, D.

1994-01-01

This Programmatic Environmental Impact Statement (PEIS) addresses the immobilization of plutonium by vitrification. Existing engineering documents, analyses, EIS, and technical publications were used and incorporated wherever possible to provide a timely response to this support effort. Although the vitrification technology is proven for the immobilization of high-level radioactive waste, more study and technical detail will be necessary to provide a comprehensive EIS that fully addresses all aspects of introduction of plutonium to the vitrification process. This document describes the concept(s) of plutonium processing as it relates to the upgrade of the DWPF and is therefore conceptual in nature. These concepts are based on technical data and experience at the Savannah River Site and will be detailed and finalized to support execution of this immobilization option

Seismic design ampersand analysis considerations for high level nuclear waste repositories

International Nuclear Information System (INIS)

Hossain, Q.A.

1993-01-01

A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid open-quotes deterministicclose quotes and open-quotes probabilisticclose quotes concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

Seismic design and analysis considerations for high level nuclear waste repositories

International Nuclear Information System (INIS)

Hossain, Q.A.

1993-01-01

A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid ''deterministic'' and ''probabilistic'' concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

Disposal of high level and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Flowers, R.H.

1991-01-01

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

A simple gold nanoparticle-mediated immobilization method to fabricate highly homogeneous DNA microarrays having higher capacities than those prepared by using conventional techniques

International Nuclear Information System (INIS)

Jung, Cheulhee; Mun, Hyo Young; Li, Taihua; Park, Hyun Gyu

2009-01-01

A simple, highly efficient immobilization method to fabricate DNA microarrays, that utilizes gold nanoparticles as the mediator, has been developed. The fabrication method begins with electrostatic attachment of amine-modified DNA to gold nanoparticles. The resulting gold-DNA complexes are immobilized on conventional amine or aldehyde functionalized glass slides. By employing gold nanoparticles as the immobilization mediator, implementation of this procedure yields highly homogeneous microarrays that have higher binding capacities than those produced by conventional methods. This outcome is due to the increased three-dimensional immobilization surface provided by the gold nanoparticles as well as the intrinsic effects of gold on emission properties. This novel immobilization strategy gives microarrays that produce more intense hybridization signals for the complementary DNA. Furthermore, the silver enhancement technique, made possible only in the case of immobilized gold nanoparticles on the microarrays, enables simple monitoring of the integrity of the immobilized DNA probe.

Current high-level waste solidification technology

International Nuclear Information System (INIS)

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

1976-01-01

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

Review: Waste-Pretreatment Technologies for Remediation of Legacy Defense Nuclear Wastes

Energy Technology Data Exchange (ETDEWEB)

Wilmarth, William R.; Lumetta, Gregg J.; Johnson, Michael E.; Poirier, Micheal R.; Thompson, Major C.; Suggs, Patricia C.; Machara, N.

2011-01-13

The U.S. Department of Energy (DOE) is responsible for retrieving, immobilizing, and disposing of radioactive waste that has been generated during the production of nuclear weapons in the United States. The vast bulk of this waste material is stored in underground tanks at the Savannah River Site in South Carolina and the Hanford Site in Washington State. The general strategy for treating the radioactive tank waste consists of first separating the waste into high-level and low-activity fractions. This initial partitioning of the waste is referred to as pretreatment. Following pretreatment, the high-level fraction will be immobilized in a glass form suitable for disposal in a geologic repository. The low-activity waste will be immobilized in a waste form suitable for disposal at the respective site. This paper provides a review of recent developments in the application of pretreatment technologies to the processing of the Hanford and Savannah River radioactive tank wastes. Included in the review are discussions of 1) solid/liquid separations methods, 2) cesium separation technologies, and 3) other separations critical to the success of the DOE tank waste remediation effort. Also included is a brief discussion of the different requirements and circumstances at the two DOE sites that have in some cases led to different choices in pretreatment technologies.

Application of magnetic nanoparticles in smart enzyme immobilization.

Science.gov (United States)

Vaghari, Hamideh; Jafarizadeh-Malmiri, Hoda; Mohammadlou, Mojgan; Berenjian, Aydin; Anarjan, Navideh; Jafari, Nahideh; Nasiri, Shahin

2016-02-01

Immobilization of enzymes enhances their properties for efficient utilization in industrial processes. Magnetic nanoparticles, due to their high surface area, large surface-to-volume ratio and easy separation under external magnetic fields, are highly valued. Significant progress has been made to develop new catalytic systems that are immobilized onto magnetic nanocarriers. This review provides an overview of recent developments in enzyme immobilization and stabilization protocols using this technology. The current applications of immobilized enzymes based on magnetic nanoparticles are summarized and future growth prospects are discussed. Recommendations are also given for areas of future research.

Immobilization of krypton in a metal matrix

International Nuclear Information System (INIS)

Whitmell, D.S.

1987-01-01

The report presents the work carried out on the immobilization of krypton in a metallic matrix by combined ion implantation and sputtering. The process has been successfully tested using 100 curies of fully active krypton in order to demonstrate that the process operates in the radiation levels which will be obtained with active gas at a reprocessing plant. A design study for a plant for fuel reprocessing has shown that the process can be simply operated, without requiring shielded cells. These results, which complete the development programme, indicate that the process is ideal for the containment of kripton arising from the processing of nuclear fuel and that the product will retain the gas under normal storage conditions and also during simulated accident conditions

Fabrication and closure development of corrosion resistant containers for Nevada's Yucca Mountain high-level nuclear waste repository

International Nuclear Information System (INIS)

Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

1989-11-01

US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 2 figs., 4 tabs

EIS Data Call Report: Plutonium immobilization plant using ceramic in new facilities at the Savannah River Site

International Nuclear Information System (INIS)

DiSabatino, A.

1998-01-01

The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors. The ceramic immobilization alternative presented in this report consists of first converting the surplus material to an oxide, followed by incorporating the plutonium oxide into a titanate-based ceramic material that is placed in metal cans

Bridging nuclear safety, security and safeguards at geological disposl of high level radioactive waste and spent nuclear fuel

International Nuclear Information System (INIS)

Niemeyer, Irmgard; Deissmann, Guido; Bosbach, Dirk

2016-01-01

Findings and recommendations: • Further R&D needed to identify concepts, methods and technologies that would be best suited for the holistic consideration of safety, security and safeguards provisions of geological disposal. • 3S ‘toolbox’, including concepts, methods and technologies for: ■ material accountancy, ■ measurement techniques for spent fuel verification, ■ containment and surveillance, ■ analysis of open source information, ■ environmental sampling and monitoring, ■ continuity of knowledge, ■ design implications. •: Bridging safety, security and safeguards in research funding and research activities related to geological disposal of high-level radioactive waste and spent nuclear fuel.

Department of Energy pretreatment of high-level and low-level wastes

International Nuclear Information System (INIS)

McGinnis, C.P.; Hunt, R.D.

1995-01-01

The remediation of the 1 x 10 8 gal of highly radioactive waste in the underground storage tanks (USTs) at five US Department of Energy (DOE) sites is one of DOE's greatest challenges. Therefore, the DOE Office of Environmental Management has created the Tank Focus Area (TFA) to manage an integrated technology development program that results in the safe and efficient remediation of UST waste. The TFA has divided its efforts into five areas, which are safety, characterization, retrieval/closure, pretreatment, and immobilization. All DOE pretreatment activities are integrated by the Pretreatment Technical Integration Manager of the TFA. For FY 1996, the 14 pretreatment tasks are divided into 3 systems: supernate separations, sludge treatment, and solid/liquid separation. The plans and recent results of these TFA tasks, which include two 25,000-gal demonstrations and two former TFA tasks on Cs removal, are presented. The pretreatment goals are to minimize the volume of high-level waste and the radioactivity in low-level waste

Nuclear waste management. Quarterly progress report, April-June 1981

Energy Technology Data Exchange (ETDEWEB)

Chikalla, T.D.; Powell, J.A.

1981-09-01

Reports and summaries are presented for the following: high-level waste process development; alternative waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; and analysis of spent fuel policy implementation.

Nuclear Waste Disposal and Strategies for Predicting Long-Term Performance of Material

International Nuclear Information System (INIS)

Wicks, G.G.

2001-01-01

Ceramics have been an important part of the nuclear community for many years. On December 2, 1942, an historic event occurred under the West Stands of Stagg Field, at the University of Chicago. Man initiated his first self-sustaining nuclear chain reaction and controlled it. The impact of this event on civilization is considered by many as monumental and compared by some to other significant events in history, such as the invention of the steam engine and the manufacturing of the first automobile. Making this event possible and the successful operation of this first man-made nuclear reactor, was the use of forty tons of UO2. The use of natural or enriched UO2 is still used today as a nuclear fuel in many nuclear power plants operating world-wide. Other ceramic materials, such as 238Pu, are used for other important purposes, such as ceramic fuels for space exploration to provide electrical power to operate instruments on board spacecrafts. Radioisotopic Thermoelectric Generators (RTGs) are used to supply electrical power and consist of a nuclear heat source and converter to transform heat energy from radioactive decay into electrical power, thus providing reliable and relatively uniform power over the very long lifetime of a mission. These sources have been used in the Galileo spacecraft orbiting Jupiter and for scientific investigations of Saturn with the Cassini spacecraft. Still another very important series of applications using the unique properties of ceramics in the nuclear field, are as immobilization matrices for management of some of the most hazardous wastes known to man. For example, in long-term management of radioactive and hazardous wastes, glass matrices are currently in production immobilizing high-level radioactive materials, and cementious forms have also been produced to incorporate low level wastes. Also, as part of nuclear disarmament activities, assemblages of crystalline phases are being developed for immobilizing weapons grade plutonium, to

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

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

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

Capture and immobilization of krypton-85

International Nuclear Information System (INIS)

Whitmell, D.S.; Geens, L.; Penzhorn, R.D.; Smith, M.J.S.

1985-01-01

It may become necessary to contain the krypton-85 released from nuclear fuel during reprocessing in order to reduce the exposure to the local population and the radioactive background throughout the world. A brief description is given of studies being carried out in the Indirect Action Programme. The separation of krypton from other off-gases by cryogenic distillation in the presence of oxygen is being studied at SCK/CEN Mol, together with the behavior of ozone in the distillation column. Two processes for the immobilization of krypton in solid forms have been successfully developed and demonstrated. At KfK Karlsruhe, krypton is encapsulated in vitrified zeolites; at AERE Harwell, krypton is immobilized within a metallic matrix. These processes offer excellent gas retention and either could be adopted for a reprocessing plant

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

Nuclear level density parameter 's dependence on angular momentum

International Nuclear Information System (INIS)

Aggarwal, Mamta; Kailas, S.

2009-01-01

Nuclear level densities represent a very important ingredient in the statistical Model calculations of nuclear reaction cross sections and help to understand the microscopic features of the excited nuclei. Most of the earlier experimental nuclear level density measurements are confined to low excitation energy and low spin region. A recent experimental investigation of nuclear level densities in high excitation energy and angular momentum domain with some interesting results on inverse level density parameter's dependence on angular momentum in the region around Z=50 has motivated us to study and analyse these experimental results in a microscopic theoretical framework. In the experiment, heavy ion fusion reactions are used to populate the excited and rotating nuclei and measured the α particle evaporation spectra in coincidence with ray multiplicity. Residual nuclei are in the range of Z R 48-55 with excitation energy range 30 to 40 MeV and angular momentum in 10 to 25. The inverse level density parameter K is found to be in the range of 9.0 - 10.5 with some exceptions

Iron phosphate glass: a promising matrix for the immobilization of Cs and Mo

International Nuclear Information System (INIS)

Hemadevi, V.; Joseph, Kitheri

2015-01-01

Presently, borosilicate glass (BSG) is the acceptable vitrification matrix for the immobilization of high level waste. The solubility of Mo in BSG is limited in the presence of Cs. As per the literature, solubility of Mo in BSG is about 2.5 wt. % in the presence of Cs. Hence it is difficult to immobilize nuclear waste rich in Cs and Mo in borosilicate glass. It is observed that the composition of Cs and Mo expressed as oxides are 10.4 and 14.7 wt. % respectively in simulated fast reactor waste. Iron phosphate glass containing 20 wt. % simulated fast reactor waste (referred as IP20FRW) was synthesized and characterized. IP20FRW contains ~ 3 wt. % of molybdenum oxide along with 2 wt. % cesium oxide. IPG is a suitable matrix for the immobilization of Cs and Mo separately. Hence it is essential to understand the glass characteristics of IPG containing both Cs and Mo. This paper explores systematic loading of both Cs and Mo such that the final composition corresponds to 10.5 wt. % Cs 2 O-15 wt. % MoO 3 -31.9 wt. % Fe 2 O 3 -42.6 wt. % P 2 O 5 . In addition to synthesis, the present study also includes understanding the change in glass characteristics of IPG containing both Cs and Mo. The possibility of higher percent loading of both Cs and Mo in IPG demonstrates the better glass forming characteristics of IPG. The synthesis and characterization of Cs-Mo loaded glasses will be discussed in this paper. (author)

Synroc - a multiphase ceramic for high level nuclear waste immobilisation

International Nuclear Information System (INIS)

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

1992-01-01

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

Production of Biodiesel Using Immobilized Lipase and the Characterization of Different Co-Immobilizing Agents and Immobilization Methods

Directory of Open Access Journals (Sweden)

Kang Zhao

2016-08-01

Full Text Available Lipase from Candida sp. 99–125 is widely employed to catalyzed transesterification and can be used for biodiesel production. In this study, the lipase was immobilized by combined adsorption and entrapment to catalyze biodiesel production from waste cooking oil (WCO via transesterification, and investigating co-immobilizing agents as additives according to the enzyme activity. The addition of the mixed co-immobilizing agents has positive effects on the activities of the immobilized lipase. Three different immobilizing methods were compared by the conversion ratio of biodiesel and structured by Atom Force Microscopy (AFM and Scanning Electron Microscopy (SEM, respectively. It was found that entrapment followed by adsorption was the best method. The effect of the co-immobilizing agent amount, lipase dosage, water content, and reuse ability of the immobilized lipase was investigated. By comparison with previous research, this immobilized lipase showed good reuse ability: the conversion ratio excesses 70% after 10 subsequent reactions, in particular, was better than Novozym435 and TLIM on waste cooking oil for one unit of lipase.

Acute immobilization stress following contextual fear conditioning reduces fear memory: timing is essential.

Science.gov (United States)

Uwaya, Akemi; Lee, Hyunjin; Park, Jonghyuk; Lee, Hosung; Muto, Junko; Nakajima, Sanae; Ohta, Shigeo; Mikami, Toshio

2016-02-24

Histone acetylation is regulated in response to stress and plays an important role in learning and memory. Chronic stress is known to deteriorate cognition, whereas acute stress facilitates memory formation. However, whether acute stress facilitates memory formation when it is applied after fear stimulation is not yet known. Therefore, this study aimed to investigate the effect of acute stress applied after fear training on memory formation, mRNA expression of brain-derived neurotrophic factor (BDNF), epigenetic regulation of BDNF expression, and corticosterone level in mice in vivo. Mice were subjected to acute immobilization stress for 30 min at 60 or 90 min after contextual fear conditioning training, and acetylation of histone 3 at lysine 14 (H3K14) and level of corticosterone were measured using western blot analysis and enzyme-linked immunosorbent assay (ELISA), respectively. A freezing behavior test was performed 24 h after training, and mRNA expression of BDNF was measured using real-time polymerase chain reactions. Different groups of mice were used for each test. Freezing behavior significantly decreased with the down-regulation of BDNF mRNA expression caused by acute immobilization stress at 60 min after fear conditioning training owing to the reduction of H3K14 acetylation. However, BDNF mRNA expression and H3K14 acetylation were not reduced in animals subjected to immobilization stress at 90 min after the training. Further, the corticosterone level was significantly high in mice subjected to immobilization stress at 60 min after the training. Acute immobilization stress for 30 min at 60 min after fear conditioning training impaired memory formation and reduced BDNF mRNA expression and H3K14 acetylation in the hippocampus of mice owing to the high level of corticosterone.

Environmental program planning for the proposed high-level nuclear waste repository at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

1987-08-01

This report was prepared to illustrate the policy and actions that the State of Nevada believe are required to assure that the quality of the environment is adequately considered during the course of the DOE work at the proposed high-level nuclear waste repository at Yucca Mountain. The report describes the DOE environmental program and the studies planned by NWPO to reflect the State's position toward environmental protection. 41 refs., 2 figs., 11 tabs

Disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Glasby, G.P.

1977-01-01

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

Other-than-high-level waste

International Nuclear Information System (INIS)

Bray, G.R.

1976-01-01

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

Structural and Chemical Characterization of Silica Spheres before and after Modification by Silanization for Trypsin Immobilization

Directory of Open Access Journals (Sweden)

Eduardo F. Barbosa

2017-01-01

Full Text Available In the last decades, silica particles of a variety of sizes and shapes have been characterized and chemically modified for several applications, from chromatographic separation to dental supplies. The present study proposes the use of aminopropyl triethoxysilane (APTS silanized silica particles to immobilize the proteolytic enzyme trypsin for the development of a bioreactor. The major advantage of the process is that it enables the polypeptides hydrolysis interruption simply by removing the silica particles from the reaction bottle. Silanized silica surfaces showed significant morphological changes at micro- and nanoscale level. Chemical characterization showed changes in elemental composition, chemical environment, and thermal degradation. Their application as supports for trypsin immobilization showed high immobilization efficiency at reduced immobilization times, combined with more acidic conditions. Indirect immobilization quantification by reversed-phase ultrafast high performance liquid chromatography proved to be a suitable approach due to its high linearity and sensitivity. Immobilized trypsin activities on nonmodified and silanized silica showed promising features (e.g., selective hydrolysis for applications in proteins/peptides primary structure elucidation for proteomics. Silanized silica system produced some preferential targeting peptides, probably due to the hydrophobicity of the nanoenvironment conditioned by silanization.

Immobilization of radioactive waste in glass matrices

International Nuclear Information System (INIS)

Wicks, G.G.

1978-01-01

A promising process for long-term management of high-level radioactive waste is to immobilize the waste in a borosilicate glass matrix. Among the most important criteria characterizing the integrity of the large-scale glass-waste forms are that they possess good chemical stability (including low leachability), thermal stability, mechanical integrity, and high radiation stability. Fulfillment of these criteria ensures the maximum margin of safety of glass-waste products, following solidification, handling, transportation, and long-term storage

High-level radioactive waste repositories site selection plan

International Nuclear Information System (INIS)

Castanon, A.; Recreo, F.

1985-01-01

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

Supplemental Immobilization Cast Stone Technology Development and Waste Form Qualification Testing Plan

Energy Technology Data Exchange (ETDEWEB)

Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R. Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, Eric M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chung, Chul-Woo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Swanberg, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2013-05-31

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrify all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.

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

International Nuclear Information System (INIS)

Cornman, W.R.

1981-08-01

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

Natural geochemical analogues of the near field of high-level nuclear waste repositories

International Nuclear Information System (INIS)

Apps, J.A.

1995-01-01

United States practice has been to design high-level nuclear waste (HLW) geological repositories with waste densities sufficiently high that repository temperatures surrounding the waste will exceed 100 degrees C and could reach 250 degrees C. Basalt and devitrified vitroclastic tuff are among the host rocks considered for waste emplacement. Near-field repository thermal behavior and chemical alteration in such rocks is expected to be similar to that observed in many geothermal systems. Therefore, the predictive modeling required for performance assessment studies of the near field could be validated and calibrated using geothermal systems as natural analogues. Examples are given which demonstrate the need for refinement of the thermodynamic databases used in geochemical modeling of near-field natural analogues and the extent to which present models can predict conditions in geothermal fields

Overview: Defense high-level waste technology program

International Nuclear Information System (INIS)

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

1987-01-01

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

R ampersand D plan for immobilization technologies: fissile materials disposition program. Revision 1.0

International Nuclear Information System (INIS)

Shaw, H.F.; Armantrout, G.A.

1996-09-01

In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long- term fissile material management options, the Department of Energy's Fissile Materials Disposition Program (FMDP) is conducting studies of options for the storage and disposition of surplus plutonium (Pu). One set of alternatives for disposition involve immobilization. The immobilization alternatives provide for fixing surplus fissile materials in a host matrix in order to create a solid disposal form that is nuclear criticality-safe, proliferation-resistant and environmentally acceptable for long-term storage or disposal

Crystallization behavior of nuclear waste forms

International Nuclear Information System (INIS)

Rusin, J.M.; Lokken, R.O.; May, R.P.; Wald, J.W.

1981-09-01

Several waste form options have been or are being developed for the immobilization of high-level wastes. The final selection of a waste form must take into consideration both waste form product as well as process factors. Crystallization behavior has an important role in nuclear waste form technology. For glass or vitreous waste forms, crystallization is generally controlled to a minimum by appropriate glass formulation and heat treatment schedules. With glass ceramic waste forms, crystallization is essential to convert glass products to highly crystalline waste forms with a minimum residual glass content. In the case of ceramic waste forms, additives and controlled sintering schedules are used to contain the radionuclides in specific tailored crystalline phases

Computer-based control of nuclear power information systems at international level

International Nuclear Information System (INIS)

Boniface, Ekechukwu; Okonkwo, Obi

2011-01-01

In most highly industrialized countries of the world information plays major role in anti-nuclear campaign. Information and discussions on nuclear power need critical and objective analysis before the structured information presentation to the public to avoid bias anti-nuclear information on one side and neglect of great risk in nuclear power. This research is developing a computer-based information system for the control of nuclear power at international level. The system is to provide easy and fast information highways for the followings: (1) Low Regulatory dose and activity limit as level of high danger for individuals and public. (2) Provision of relevant technical or scientific education among the information carriers in the nuclear power countries. The research is on fact oriented investigation about radioactivity. It also deals with fact oriented education about nuclear accidents and safety. A standard procedure for dissemination of latest findings using technical and scientific experts in nuclear technology is developed. The information highway clearly analyzes the factual information about radiation risk and nuclear energy. Radiation cannot be removed from our environment. The necessity of radiation utilizations defines nuclear energy as two-edge sword. It is therefore, possible to use computer-based information system in projecting and dissemination of expert knowledge about nuclear technology positively and also to use it in directing the public on the safety and control of the nuclear energy. The computer-based information highway for nuclear energy technology is to assist in scientific research and technological development at international level. (author)

Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 1, Discussion and glass durability data

International Nuclear Information System (INIS)

Kindle, C.H.; Kreiter, M.R.

1987-12-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. Data collected throughout the world are included in the data base; current emphasis is on waste glasses and their properties. The goal is to provide a data base of properties and compositions and an analysis of dominant property trends as a function of composition. This data base is a resource that nuclear waste producers, disposers, and regulators can use to compare properties of a particular high-level nuclear waste glass product with the properties of other glasses of similar compositions. Researchers may use the data base to guide experimental tests to fill gaps in the available knowledge or to refine empirical models. The data are incorporated into a computerized data base that will allow the data to be extracted based on, for example, glass composition or test duration. 3 figs

Biodiesel production with immobilized lipase: A review.

Science.gov (United States)

Tan, Tianwei; Lu, Jike; Nie, Kaili; Deng, Li; Wang, Fang

2010-01-01

Fatty acid alkyl esters, also called biodiesel, are environmentally friendly and show great potential as an alternative liquid fuel. Biodiesel is produced by transesterification of oils or fats with chemical catalysts or lipase. Immobilized lipase as the biocatalyst draws high attention because that process is "greener". This article reviews the current status of biodiesel production with immobilized lipase, including various lipases, immobilization methods, various feedstocks, lipase inactivation caused by short chain alcohols and large scale industrialization. Adsorption is still the most widely employed method for lipase immobilization. There are two kinds of lipase used most frequently especially for large scale industrialization. One is Candida antartica lipase immobilized on acrylic resin, and the other is Candida sp. 99-125 lipase immobilized on inexpensive textile membranes. However, to further reduce the cost of biodiesel production, new immobilization techniques with higher activity and stability still need to be explored. Copyright 2010 Elsevier Inc. All rights reserved.

Overall review strategy for the Nuclear Regulatory Commission's High-Level Waste Repository Program

International Nuclear Information System (INIS)

Johnson, R.L.

1994-11-01

The Overall Review Strategy gives general guidance to the Nuclear Regulatory Commission staff for conducting it's license application and pre-license application reviews. These reviews are in support of the Commission's construction authorization decision for a geologic repository for the disposal of high-level radioactive waste. Objectives and strategies are defined that focus the staff's reviews on determining compliance with requirements of 10 CFR Part 60. These strategies define how the staff prioritizes its reviews on those key technical uncertainties considered to be most important to repository performance. Strategies also give guidance for developing, in an integrated way, the License Application Review Plan together with supporting performance assessments, analyses, and research

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

Hoza, M.

1996-01-01

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

Metabolic alkalosis during immobilization in monkeys (M. nemestrina)

Science.gov (United States)

Young, D. R.; Yeh, I.; Swenson, R. S.

1983-01-01

The systemic and renal acid-base response of monkeys during ten weeks of immobilization was studied. By three weeks of immobilization, arterial pH and bicarbonate concentrations were elevated (chronic metabolic alkalosis). Net urinary acid excretion increased in immobilized animals. Urinary bicarbonate excretion decreased during the first three weeks of immobilization, and then returned to control levels. Sustained increases in urinary ammonium excretion were seen throughout the time duration of immobilization. Neither potassium depletion nor hypokalemia was observed. Most parameters returned promptly to the normal range during the first week of recovery. Factors tentatively associated with changes in acid-base status of monkeys include contraction of extracellular fluid volume, retention of bicarbonate, increased acid excretion, and possible participation of extrarenal buffers.

High-level waste processing at the Savannah River Site: An update

International Nuclear Information System (INIS)

Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

1997-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ''sludge-only'' composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ''coupled'' feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates

Peer review for high-level nuclear waste repositories: Generic technical position

International Nuclear Information System (INIS)

Altman, W.D.; Donnelly, J.P.; Kennedy, J.E.

1988-02-01

This document provides guidance on the use of the peer review process in the high-level nuclear waste repository program. The applicant must demonstrate in the license application that the applicable health, safety, and environmental regulations in 10 CFR Part 60 have been met. Confidence in the data used to support the license application is obtained through a quality assurance (AQ) program. Peer reviews may be used as part of the QA actions necessary to provide adequate confidence in the work being reviewed. Because of several unique conditions inherent to the geologic repository program, expert judgment will need to be utilized in assessing the adequacy of work. Peer reviews are a mechanism by which these judgments may be made. This document provides guidance on areas where a peer review is appropriate, the acceptability of peers, and the conduct and documentation of a peer review

Solidification of Savannah River Plant high-level waste

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1983-11-01

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

Public opposition to the siting of the high-level nuclear waste repository: The importance of trust

International Nuclear Information System (INIS)

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

1991-01-01

This paper examines several dimensions of public opposition to the proposed siting of the high-level nuclear waste repository at Yucca Mountain. In order to provide a context for the public's views of the repository in metropolitan Clark County, both governmental studies of the repository siting process are analyzed, as well as elements of the Nuclear Waste Policy Act. This analysis suggests that one potentially key component of the public's opposition to the siting, as well as their perceptions of risk of the facility, may be the result of a lack of trust in the Department of Energy. Empirical analysis of survey data collected in Nevada in 1988 confirms the strong relationship between political trust and repository risk perceptions

Idaho Nuclear Technology and Engineering Center Newly Generated Liquid Waste Demonstration Project Feasibility Study

International Nuclear Information System (INIS)

Herbst, A.K.

2000-01-01

A research, development, and demonstration project for the grouting of newly generated liquid waste (NGLW) at the Idaho Nuclear Technology and Engineering Center is considered feasible. NGLW is expected from process equipment waste, decontamination waste, analytical laboratory waste, fuel storage basin waste water, and high-level liquid waste evaporator condensate. The potential grouted waste would be classed as mixed low-level waste, stabilized and immobilized to meet RCRA LDR disposal in a grouting process in the CPP-604 facility, and then transported to the state

Attitudes and opposition in siting a high level nuclear waste repository

International Nuclear Information System (INIS)

Sjoeberg, L.; Viklund, M.; Truedsson, J.

1998-09-01

In Sweden, the Swedish Nuclear Fuel and Waste Management Company (SKB) handles all issues concerning nuclear waste, including the siting process, in which the final outcome is intended to be a repository for high level nuclear waste placed deep down in bedrock. The main objective of the siting process is to find a host community fulfilling two important conditions: the safety demands have been met and agreements with the municipality can be accomplished. Only in such municipalities, so-called feasibility studies will be conducted. After conducting general studies in the whole country, SKB, in October 1992, sent letters with information about the intended feasibility studies to all Swedish municipalities. As a result, feasibility studies are or have been considered - and in some cases also been conducted - in eleven Swedish municipalities up until 1998. These are the municipalities where the attitudes and opposition towards a feasibility study, and possibly a final repository, are studied. The discussion can be divided into three main parts: Management of the siting process; Inherent 'chaotic' processes and/or factors and risk perception. It is argued that two important problems could have been avoided at least partly: The citizens in many municipalities were uncertain of the relationship between a feasibility study and a final repository, and in many municipalities the citizens were afraid that the Government could overrule the municipal veto. Because of these fears, a common argument among the opponents of a feasibility study was: 'to be sure of not receiving a final repository, we say no to a feasibility study'. Some inherent factors, more or less prevalent in the municipalities as well as in society in general, may also partly explain the outcome of the siting process. The municipalities in which the debate has been heated, and where public support has been more difficult to reach, share some common characteristics. Esp. in the municipalities in the north of

Attitudes and opposition in siting a high level nuclear waste repository

Energy Technology Data Exchange (ETDEWEB)

Sjoeberg, L.; Viklund, M.; Truedsson, J

1998-09-01

In Sweden, the Swedish Nuclear Fuel and Waste Management Company (SKB) handles all issues concerning nuclear waste, including the siting process, in which the final outcome is intended to be a repository for high level nuclear waste placed deep down in bedrock. The main objective of the siting process is to find a host community fulfilling two important conditions: the safety demands have been met and agreements with the municipality can be accomplished. Only in such municipalities, so-called feasibility studies will be conducted. After conducting general studies in the whole country, SKB, in October 1992, sent letters with information about the intended feasibility studies to all Swedish municipalities. As a result, feasibility studies are or have been considered - and in some cases also been conducted - in eleven Swedish municipalities up until 1998. These are the municipalities where the attitudes and opposition towards a feasibility study, and possibly a final repository, are studied. The discussion can be divided into three main parts: Management of the siting process; Inherent `chaotic` processes and/or factors and risk perception. It is argued that two important problems could have been avoided at least partly: The citizens in many municipalities were uncertain of the relationship between a feasibility study and a final repository, and in many municipalities the citizens were afraid that the Government could overrule the municipal veto. Because of these fears, a common argument among the opponents of a feasibility study was: `to be sure of not receiving a final repository, we say no to a feasibility study`. Some inherent factors, more or less prevalent in the municipalities as well as in society in general, may also partly explain the outcome of the siting process. The municipalities in which the debate has been heated, and where public support has been more difficult to reach, share some common characteristics. Esp. in the municipalities in the north of

[Native, modified, and immobilized chymotrypsin in chaotropic media. Stabilization limits].

Science.gov (United States)

Panova, A A; Levitskiĭ, V Iu; Mozhaev, V V

1994-07-01

To stabilize alpha-chymotrypsin against irreversible thermal inactivation at high temperatures, methods of covalent modification and multi-point immobilization in combination with the addition of salting-in compounds were used. The upper limit of the protein stability proved to be the same for a combination of the modification and salting-in media and for each of these methods separately. The limit of stabilization reached by means of covalent immobilization is higher than the limit of stabilization reached by two other methods. The greatest stabilization of immobilized alpha-chymotrypsin by the salting-in media (a 10000 fold increase in the native enzyme's stability level) takes place only in the case of the protein with the minimum number of bonds with the support. Stabilization of the enzyme by these methods is explained in terms of the suppression of the conformational inactivation processes.

Study of nuclear level densities for exotic nuclei

International Nuclear Information System (INIS)

Nasrabadi, M. N.; Sepiani, M.

2012-01-01

Nuclear level density is one of the properties of nuclei with widespread applications in astrophysics and nuclear medicine. Since there has been little experimental and theoretical research on the study of nuclei which are far from stability line, studying nuclear level density for these nuclei is of crucial importance. Also, as nuclear level density is an important input for nuclear research codes, hence studying the methods for calculation of this parameter is essential. Besides introducing various methods and models for calculating nuclear level density for practical applications, we used exact spectra distribution (SPDM) for determining nuclear level density of two neutron and proton enriched exotic nuclei with the same mass number.

Stabilization and immobilization of military plutonium: A non-proliferation perspective

Energy Technology Data Exchange (ETDEWEB)

Leventhal, P. [Nuclear Control Institute, Washington, DC (United States)

1996-05-01

The Nuclear Control Institute welcomes this DOE-sponsored technical workshop on stabilization and immobilization of weapons plutonium (W Pu) because of the significant contribution it can make toward the ultimate non-proliferation objective of eliminating weapons-usable nuclear material, plutonium and highly enriched uranium (HEU), from world commerce. The risk of theft or diversion of these materials warrants concern, as only a few kilograms in the hands of terrorists or threshold states would give them the capability to build nuclear weapons. Military plutonium disposition questions cannot be addressed in isolation from civilian plutonium issues. The National Academy of Sciences has urged that {open_quotes}further steps should be taken to reduce the proliferation risks posed by all of the world`s plutonium stocks, military and civilian, separated and unseparated...{close_quotes}. This report discusses vitrification and a mixed oxide fuels option, and the effects of disposition choices on civilian plutonium fuel cycles.

Diesel oil removal by immobilized Pseudoxanthomonas sp. RN402.

Science.gov (United States)

Nopcharoenkul, Wannarak; Netsakulnee, Parichat; Pinyakong, Onruthai

2013-06-01

Pseudoxanthomonas sp. RN402 was capable of degrading diesel, crude oil, n-tetradecane and n-hexadecane. The RN402 cells were immobilized on the surface of high-density polyethylene plastic pellets at a maximum cell density of 10(8) most probable number (MPN) g(-1) of plastic pellets. The immobilized cells not only showed a higher efficacy of diesel oil removal than free cells but could also degrade higher concentrations of diesel oil. The rate of diesel oil removal by immobilized RN402 cells in liquid culture was 1,050 mg l(-1) day(-1). Moreover, the immobilized cells could maintain high efficacy and viability throughout 70 cycles of bioremedial treatment of diesel-contaminated water. The stability of diesel oil degradation in the immobilized cells resulted from the ability of living RN402 cells to attach to material surfaces by biofilm formation, as was shown by CLSM imaging. These characteristics of the immobilized RN402 cells, including high degradative efficacy, stability and flotation, make them suitable for the purpose of continuous wastewater bioremediation.

Siting the nation's first high-level nuclear waste repository: Social impacts for Utah

International Nuclear Information System (INIS)

Olshansky, S.J.

1986-01-01

The siting of a high-level nuclear waste repository in the United States has been an issue of great controversy, particulary for the states under consideration. In addition to concerns expressed about the geological stability of the proposed sites, numerous social issues have been raised by the general public--most of which have been addressed by the draft environmental impact statements. Among the social impacts raised by the Department of Engery and the general public, those receiving the greatest attention were the potential influence of the repository on local economics, tourism, and the health status of the local residents. The major issues of interest in the present study include 1) the effects of respondent knowledge of nuclear waste disposal issues on opinions of health effects and tourism, particularly as they are affected by visitation patterns, and 2) the effects of occupation and education (in particular) on knowledge of nuclear waste disposal issues and opinions on technical and non-technical aspects of siting the repository. Preliminary results indicate that only about 40 percent of the respondents have visited the national parks in southeastern Utah, but over 70 percent feel they are informed about the issues associated with siting the repository. Over 60 percent of the respondents were very concerned about the possible negative effect the repository could have on jobs, tourism, health effects, and environmental quality. Cross-tabulations indicate that the respondents self rating on knowledge of nuclear disposal issues has a statistically significant influence on responses to socioeconomic issues, yet the same self rating scale is significantly influenced by the frequency with which respondents have visited the national parks in southeastern Utah

Supply and cost factors for metals in the Canadian nuclear fuel waste immobilization program

International Nuclear Information System (INIS)

McConnell, D.B.

1982-11-01

Estimates have been made of the demand for immobilization containers to accommodate the irradiated fuel bundles arising from Canadian nuclear generating stations to the year 2020. The resulting estimates for container shells and container-filling alloys were compared to estimates for Canadian and Western World production of the candiate metals. The results indicate that, among the container shell metals, supply difficulties might arise only for Grade 7 titanium. Among the filling metals, only lead-antimony alloy might present supply problems. Current cost figures for plate made of each shell metal, and bulk quantities of filling metals, were compared. Materials costs would be least for a supported shell of stainless steel, followed by copper, titanium alloys Grades 2, 12 and 7, and Inconel 625. Aluminum-silicon is the lowest-cost filling matrix, followed by zinc, lead, and lead-antimony. Container durability, vault conditions, groundwater composition and other factors may play an overriding role in the final selection of materials for container construction

Compound immobilization and drug-affinity chromatography.

Science.gov (United States)

Rix, Uwe; Gridling, Manuela; Superti-Furga, Giulio

2012-01-01

Bioactive small molecules act through modulating a yet unpredictable number of targets. It is therefore of critical importance to define the cellular target proteins of a compound as an entry point to understanding its mechanism of action. Often, this can be achieved in a direct fashion by chemical proteomics. As with any affinity chromatography, immobilization of the bait to a solid support is one of the earliest and most crucial steps in the process. Interfering with structural features that are important for identification of a target protein will be detrimental to binding affinity. Also, many molecules are sensitive to heat or to certain chemicals, such as acid or base, and might be destroyed during the process of immobilization, which therefore needs to be not only efficient, but also mild. The subsequent affinity chromatography step needs to preserve molecular and conformational integrity of both bait compound and proteins in order to result in the desired specific enrichment while ensuring a high level of compatibility with downstream analysis by mass spectrometry. Thus, the right choice of detergent, buffer, and protease inhibitors is also essential. This chapter describes a widely applicable procedure for the immobilization of small molecule drugs and for drug-affinity chromatography with subsequent protein identification by mass spectrometry.

Water level control for a nuclear steam generator

International Nuclear Information System (INIS)

Wen Tan

2011-01-01

Research highlights: → A water level control system for a nuclear steam generator (SG) is proposed. → The parameters of the control system are directly related to those of the plant model thus scheduling is easy to implement in practice. → The proposed gain-scheduled controller can achieve good performance at both low and high power levels. - Abstract: A water level control system for a nuclear steam generator (SG) is proposed. The control system consists of a feedback controller and a feedforward controller. The feedback controller is of first order, the feedforward controller is of second order, and parameters of the two controllers are directly related to the parameters of plant model thus scheduling is easy to implement in practice. Robustness and performance of the feedback and the feedforward controllers are analyzed in details and tuning of the two parameters of the controllers are discussed. Comparisons among a single robust controller, a multi-model controller and a gain-scheduled controller are studied. It is shown that the proposed gain-scheduled controller can achieve good performance at both low and high power levels.

Immobilization of Bovine Serum Albumin Upon Multiwall Carbon Nanotube for High Speed Humidity Sensing Application.

Science.gov (United States)

Bhattacharya, Sankhya; Sasmal, Milan

2016-01-01

We present a high-speed humidity sensor based on immobilization of bovine serum albumin upon multiwall carbon nanotube (IBC). A simple and versatile drop casting technique was employed to make the humidity sensor using novel material IBC at room temperature. IBC was synthesized using easy solution process technique. The working principle of the IBC humidity sensor depends upon the variation of output current or conductance with the exposure of different humidity level. Humidity sensing properties of our device is explained on the basis of charge transfer from water molecules to IBC and bovine serum albumin to multiwall carbon nanotube (MWCNT). Our sensor exhibits faster response time around 1.2 s and recovery time 1.5 s respectively.

MUCOADHESIVE GEL WITH IMMOBILIZED LYSOZYME: PREPARATION AND PROPERTIES

Directory of Open Access Journals (Sweden)

Dekina S. S.

2015-08-01

Full Text Available The study of non-covalent immobilized lysozyme, as well as physico-chemical and biochemical properties of obtained mucoadhesive gel was the aim of the research. Lysozyme activity was determined by bacteriolytic method (Micrococcus lysodeikticus cells acetone powder was a substrate. Lysozyme immobilization was conducted by the method of entrapment in gel. Enzyme carrier interaction was studied by viscometric, spectrophotometric and spectrofluorimetric methods. Mucoadhesive gel with immobilized lysozyme, possessing antiinflammatory and antimicrobial activities, was prepared. Due to immobilization, protein-polymer complex with the original enzymatic activity was formed. The product is characterized by high mucoadhesive properties, quantitative retaining of protein and bacteriolytic activity, prolonged release of the enzyme, improved biochemical characteristics (extended pH-activity profile, stability in acidic medium and during storage for 2 years, and it is perspective for further studies. The proposed method for lysozyme immobilization in the carboxymethyl cellulose sodium salt gel allows to obtain a stable, highly efficient product, with high adhesive properties for attachment to the mucous membranes, that is promising for use in biomedicine.

Pressure ulcers from spinal immobilization in trauma patients: a systematic review.

Science.gov (United States)

Ham, Wietske; Schoonhoven, Lisette; Schuurmans, Marieke J; Leenen, Luke P H

2014-04-01

To protect the (possibly) injured spine, trauma patients are immobilized on backboard or vacuum mattress, with a cervical collar, lateral headblocks, and straps. Several studies identified pressure ulcer (PU) development from these devices. The aim of this literature study was to gain insight into the occurrence and development of PUs, the risk factors, and the possible interventions to prevent PUs related to spinal immobilization with devices in adult trauma patients. We systematically searched PubMed (MEDLINE), EMBASE, Cochrane, and CINAHL for the period 1970 to September 2011. Studies were included if participants were healthy volunteers under spinal immobilization or trauma patients under spinal immobilization until spine injuries were diagnosed or excluded. Outcomes of primary interest included occurrence, severity, and risk for PU development as well as prevention of PU development related to spinal immobilization devices. The results of included studies show an incidence of collar-related PUs ranging from 6.8% to 38%. Described locations are the occiput, chin, shoulders, and back. The severity of these PUs varies between Stages 1 and 3, and one study describes PUs requiring surgical debridement, indicating a Stage 4 PU. Described risk factors for PU development are high pressure and pain from immobilizing devices, the length of time in/on a device, intensive care unit admission, high Injury Severity Scores (ISSs), mechanical ventilation, and intracranial pressure monitoring. Preventive interventions for collar-related PUs include early replacement of the extrication collar and regular skin assessment, collar refit, and position change. The results from this systematic review show that immobilization with devices increases the risk for PU development. This risk is demonstrated in nine experimental studies with healthy volunteers and in four clinical studies. Systematic review, level III.

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

International Nuclear Information System (INIS)

Radulesscu, G.; Tang, J.S.

2000-01-01

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M andO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M andO 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M andQ 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M andO 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

Energy Technology Data Exchange (ETDEWEB)

G. Radulesscu; J.S. Tang

2000-06-07

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable

Technetium Chemistry in High-Level Waste

International Nuclear Information System (INIS)

Hess, Nancy J.

2006-01-01

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

International high-level radioactive waste repositories

International Nuclear Information System (INIS)

Lin, W.

1996-01-01

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

Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

Science.gov (United States)

Pang, Bo; Becker, Frank

2017-04-28

Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Microwave process employed to study the immobilization feasibility of spent ion exchange resins in polymeric matrices

Energy Technology Data Exchange (ETDEWEB)

Caratin, Reinaldo L.; Araujo, Sumair G. de; Landini, Liliane; Neves, Sabrina C.; Lugao, Ademar B. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], E-mail: rcaratin@ipen.br, E-mail: sgaraujo@ipen.br, E-mail: llandini@ipen.br, E-mail: scneves@ipen.br, E-mail: ablugao@ipen.br

2007-07-01

Nuclear activities generate radioactive wastes in several physical states, radioactive levels and kinds of radioactive emission. Hence, a lot of techniques have been developed and optimized to do the immobilization of these materials, according to local and international regulations to protect human being and environment. Another great concern is the indiscriminate disposal of used polymeric materials (such as plastic and rubber) or production leftovers in landfills, which remain for many years before they are naturally decomposed. In this work, it was studied the possibility of carrying out the immobilization of spent ion exchange resins (contaminated with ionising radiation), by using polymeric matrices of bitumen and rubber (as solidification materials for the storage of low level radioactive waste). The samples were mixed at different percentages and were heated in a microwave device (2,450 MHz) at IPEN/CNEN-SP, varying the irradiation time and power. The objective of the immobilization is converting the wastes into forms that are leach resistant and physically and chemically stable for disposal. Characterizations of these materials have been performed according to ABNT-NBR standards. The results indicated the previous idea of the necessary minimum temperature to keep the matrix for future embedding of radioactive waste, in solid state. (author)

Unintentional stoppages of trucks hauling high-level nuclear waste in Nevada

International Nuclear Information System (INIS)

Glickman, T.S.

1991-01-01

Any unintentional stoppage of trucks hauling high-level nuclear waste in Nevada could intensify public concerns about the negative impacts of the Yucca Mountain Project. This paper provides estimates of the expected number and frequency of such stoppages under several alternative scenarios for the routing and volume of the anticipated truck shipments. At least twice as many truck stoppages are expected to occur on freeways as on non-freeways, and at least twice as many are expected to occur in rural areas as in other areas. The expected frequency of such stoppages in the state is estimated to range from about 3 per year to one every 3 1/2 years. About one in every 4 of the stoppages is expected to take place in Las Vegas and 3 out of every 4 are expected to be non-accidents such as mechanical disablements. The minimum duration of the stoppages is expected to be on the order of half an hour for disablements and an hour for accidents

Management of primary anterior shoulder dislocations using immobilization.

Science.gov (United States)

Smith, Brent I; Bliven, Kellie C Huxel; Morway, Genoveffa R; Hurbanek, Jason G

2015-05-01

Reference/Citation : Paterson WH, Throckmorton TW, Koester M, Azar FM, Kuhn JE. Position and duration of immobilization after primary anterior shoulder dislocation: a systemic review and meta-analysis of the literature. J Bone Joint Surg Am. 2010;92(18):2924-2933. Does an optimum duration and position of immobilization after primary anterior shoulder dislocation exist for reducing recurrence rates? MEDLINE/PubMed, EMBASE, and Cochrane databases were searched up to December 2009 without limitations. The search terms for all databases used were shoulder AND dislocation and shoulder AND immobilization. Criteria used to include articles were (1) English language, (2) prospective level I or level II studies (according to Journal of Bone & Joint Surgery guidelines), (3) nonoperative management of initial anterior shoulder dislocation, (4) minimum follow-up of 1 year, and (5) rate of recurrent dislocation as a reported outcome. A standardized evaluation method was used to extract data to allow assessment of methods issues and statistical analysis to determine sources of bias. The primary outcome was the recurrence rate after nonoperative management of anterior shoulder dislocation. Additional data extracted and used in subanalyses included duration and position of immobilization and age at the time of initial dislocation. Data were analyzed to determine associations among groups using 2-tailed Fisher exact tests. For pooled categorical data, relative risk of recurrent dislocation, 95% confidence intervals, and heterogeneity using the I(2) statistic and χ(2) tests were calculated for individual studies. The Mantel-Haenszel method was used to combine studies and estimate overall relative risk of recurrent dislocation and 95% confidence intervals. The statistical difference between duration of immobilization and position was determined using z tests for overall effect. Pooled results were presented as forest plots. In the initial search of the databases, the authors

Transmutation of high level nuclear waste in an accelerator driven system: towards a demonstration device of industrial interest (EUROTRANS)

International Nuclear Information System (INIS)

Knebel, Joachim U.; Ait Abderrahim, Hamid; Caron-Carles, Marylise

2010-01-01

The Integrated Project EUROTRANS (EURopean Research Programme for the TRANSmutation of High Level Nuclear Waste in an Accelerator Driven System) within the ongoing EURATOM 6th Framework Programme (FP6) is devoted to the study of transmutation of high-level waste from nuclear power plants. The work is focused on transmutation in an Accelerator Driven System (ADS). The objective of EUROTRANS is the assessment of the design and the feasibility of an industrial ADS prototype dedicated to transmutation. The necessary R and D results in the areas of accelerator components, fuel development, structural materials, thermal-hydraulics, heavy liquid metal technology and nuclear data will be made available, together with the experimental demonstration of the ADS component coupling. The outcome of this work will allow to provide a reasonably reliable assessment of technological feasibility and a cost estimate for ADS based transmutation, and to possibly decide on the detailed design of an experimental ADS and its construction in the future. EUROTRANS is integrating activities of 51 participants from 16 countries, within the industry (10 participants), the national research centres (20) and 17 universities. 16 universities are collectively represented by ENEN (European Nuclear Education Network). EUROTRANS is the continuation of the three FP5 Clusters FUETRA, BASTRA and TESTRA together with the PDS-XADS Project. It is a five-year project which started in April 2005

The precautionary principle and high-level nuclear waste policy

International Nuclear Information System (INIS)

Frishman, S.

1999-01-01

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

IR and Raman spectroscopy of sodium aluminophosphate vitreous materials for immobilization of high level wastes from nuclear fuel reprocessing

International Nuclear Information System (INIS)

Stefanovskij, S.V.; Myasoedov, B.F.; Remizov, M.B.; Belanova, E.A.

2014-01-01

The structure of sodium aluminophosphate vitreous materials containing constituents of high level wastes (cesium, magnesium, copper and molybdenum oxides) from uranium-graphite reactor has been studied by IR and Raman spectroscopy techniques coupled with X-ray diffraction. The structural network has been shown to be composed of short phosphorus-oxygen chains with embedded aluminum-oxygen tetrahedra. In the magnesium-bearing samples the cross-linking with Mg 2+ ions is possible. The effect of other oxides (Cs 2 O, MoO 3 , CuO) on the glass structure is negligible for the occuring amounts. The glasses have been devitrified partly at the quenching and much stronger at the annealing. It is reflected in splitting of the vibration bands of the bonds in structural units of the anionic motif of the vitreous materials. (authors)

High temperature crystallographic and thermodynamic investigations on synthetic calzirtite (Ca2Zr5Ti2O16)

International Nuclear Information System (INIS)

Jafar, M.; Phapale, S.; Achary, S.N.; Mishra, R.; Tyagi, A.K.

2016-01-01

Immobilization of actinides in the high level waste (HLW) produced from nuclear reactors in durable host matrix is one of the important concerns in nuclear power technology. Rock analogue (SYNROC) ceramic composites of titanates and zirconates namely zirconolite, calzirtite etc. have been proposed as alternate host matrix for disposal of long lived fission products. These minerals have ability to incorporate or immobilize a wider variety of ions simultaneously without further segregation to any other phases and are stable in geothermal conditions. Knowledge of thermodynamic stability of these minerals is important for their deployment as host matrix for actinide waste disposal. In this work crystal structure and thermodynamic parameters of a mineral analogous titanate termed as calzirtite (Ca 2 Zr 5 Ti 2 O 16 ) has been determined

High level nuclear waste repository in salt: Sealing systems status and planning report: Draft report

International Nuclear Information System (INIS)

1985-09-01

This report documents the initial conceptual design studies for a repository sealing system for a high-level nuclear waste repository in salt. The first step in the initial design studies was to review the current design level, termed schematic designs. This review identified practicality of construction and development of a design methodology as two key issues for the conceptual design. These two issues were then investigated during the initial design studies for seal system materials, seal placement, backfill emplacement, and a testing and monitoring plan. The results of these studies have been used to develop a program plan for completion of the sealing system conceptual design. 60 refs., 26 figs., 18 tabs

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Rechard, R.P.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency's Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories

Some legal aspects on high level radioactive waste disposal in Japan

International Nuclear Information System (INIS)

Tanabe, Tomoyuki

1997-01-01

In Japan, it is considered to be an urgent problem to prepare the system for the research and execution of high level radioactive waste disposal. Under what regulation scheme the disposal should be done has not been sufficiently examined. In this research, the examination was carried out on the legal aspects of high level radioactive waste disposal as follows. First, the current legislation on the disposal in Japan was analyzed, and it was made clear that high level radioactive waste disposal has not been stipulated clearly. Next, on the legal choices which are conceivable on the way the legislation for high level radioactive waste disposal should be, from the aspects of applying the law on regulating nuclear reactors and others, applying the law on nuclear power damage reparation, and industrialization by changing the government ordinances, those were arranged in six choices, and the examination was carried out for each choice from the viewpoints of the relation with the base stipulation for waste-burying business, the speciality of high level radioactive waste disposal as compared with other actions of nuclear power business, the coordination with existing nuclear power of nuclear power business, the coordination with existing nuclear power law system and the formation of national consensus. In this research, it was shown that the execution of high level radioactive waste disposal as the business based on the separate legislation is the realistic choice. (K.I.)

High-level radioactive waste disposal type and theoretical analyses

International Nuclear Information System (INIS)

Lu Yingfa; Wu Yanchun; Luo Xianqi; Cui Yujun

2006-01-01

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

STATE OF THE ART OF DRILLING LARGE DIAMETER BOREHOLES FOR DEPOSITION OF HIGH LEVEL WASTE AND SPENT NUCLEAR FUEL

Directory of Open Access Journals (Sweden)

Trpimir Kujundžić

2012-07-01

Full Text Available Deep geological disposal is internationally recognized as the safest and most sustainable option for the long-term management of high-level radioactive waste. Mainly, clay rock, salt rock and crystalline rock are being considered as possible host rocks. Different geological environment in different countries led to the various repository concepts. Main feature of the most matured repository concept is that canisters with spent nuclear fuel are emplaced in vertical or horizontal large diameter deposition holes. Drilling technology of the deposition holes depends on repository concept and geological and geomechanical characteristics of the rock. The deposition holes are mechanically excavated since drill & blast is not a possible method due to requirements on final geometry like surface roughness etc. Different methods of drilling large diameter boreholes for deposition of high-level waste and spent nuclear fuel are described. Comparison of methods is made considering performance and particularities in technology.

Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

International Nuclear Information System (INIS)

Forsberg, C.W.; Elam, K.R.

1995-01-01

With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

Technologies for destruction of long-lived radionuclides in high-level nuclear waste - overview and requirements

International Nuclear Information System (INIS)

Arthur, E.D.

1993-01-01

A major issue surrounding current nuclear power generation is the management and disposal of long-lived, high-level waste (HLW). The planned and scientifically acceptable destination for this waste is in deep underground, geologically stable, repositories. However, public concerns surrounding such disposal of long-lived nuclear wastes and other issues such as proliferation and safety negatively affect the potential role that nuclear power can play in meeting current and future national energy needs. This paper and this topical session on nuclear waste minimization, management, and remediation focus on two nuclear systems and their associated technologies that have the potential to address concerns surrounding long-lived radionuclides in HLW. Both systems offer technology applicable to HLW from current light water reactors (LWRs). In addition, these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes long-term fuel supplies, and enhanced safety. The first system is the integral fast reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while burning long-lived actinide waste. The IFR applications include the burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, accelerator transmutation of waste (ATW), is based on an accelerator-induced intense source of thermal neutrons and is aimed at the destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent-fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

Gamma radiolysis effects on leaching behavior of ceramic materials for nuclear fuel waste immobilization containers

International Nuclear Information System (INIS)

Onofrei, M.; Raine, D.K.; Hocking, W.H.; George, K.; Betteridge, J.S.

1986-01-01

The leaching behavior of ceramic materials for nuclear fuel waste immobilization containers, under the influence of a moderate gamma dose rate (4 Gy/h), has been investigated. Samples of Al/sub 2/O/sub 3/, stabilized ZrO/sub 2/, TiO/sub 2/, cermet (70% Al/sub 2/O-30% TiC), porcelain (with high Al/sub 2/O/sub 3/ content), and concrete (with sulfate-resisting portland cement plus silica fume) have been leached in Standard Canadian Shield Saline Solution (SCSSS), and SCSSS plus clay and sand (components of the disposal system), at 100 0 and 150 0 C for 231 and 987 days, respectively. Leaching solutions were analyzed and the surfaces of the leached samples were investigated by scanning electron microscopy in conjunction with energy dispersive X-ray spectroscopy and secondary ion mass spectrometry. Radiolysis did not appear to enhance the leaching, with or without bentonite and sand in the system. Analysis of the gas phase from sealed capsules showed O/sub 2/ depletion and production of CO/sub 2/ in all experiments containing bentonite. The decrease in O/sub 2/ is attributed to the leaching from the clay of Fe(II) species, which can participate in redox reactions with radicals generated by radiolysis. The CO/sub 2/ is produced from either the organic or inorganic fraction in the bentonite

Plutonium Immobilization Project - Robotic canister loading

International Nuclear Information System (INIS)

Hamilton, R.L.

2000-01-01

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

Collagen Type III and VI Turnover in Response to Long-Term Immobilization.

Directory of Open Access Journals (Sweden)

Shu Sun

Full Text Available Muscle mass and function are perturbed by immobilization and remobilization. When muscle mass changes, the quality and quantity of the extracellular matrix protein, particularly the collagens, change with it. In this study, we investigated the temporal profile of three peptide biomarkers derived from turnover of collagen type III and type VI in a long-term immobilization and remobilization study. We also compared individual biomarker levels with Lean body Mass (LBM and changes therein, hypothesizing that these biomarkers would be biomarkers of the remodeling processes associated with immobilization and/or remobilization.In the Berlin bed rest study, 20 young men were recruited and randomly assigned to 8-week's strict bed rest with or without resistive vibration exercise countermeasure. We measured three neo-epitope ELISA kits in the serum samples of this study: Pro-C3, measured the synthesis of collagen type III; Pro-C6, measured the synthesis of collagen type VI; and C6M measured the degradation of collagen type VI induced by MMP-2 and MMP-9 cleavage.Pro-C3 and Pro-C6 biomarkers are up-regulated with both immobilization and remobilization, whereas C6M is hardly affected at all. We found that Pro-C3 and C6M levels are related to LBM at baseline and that high levels of Pro-C6 are associated with smaller changes in muscle mass during both immobilization and remobilization.The Pro-C3 and-C6 biomarkers change likely reflect remodeling changes in response to unloading or reloading, whereas C6M does not appear to respond to unloading. Pro-C3 and C6M levels correlate with LBM at baseline, while Pro-C6 is related to the anabolic and catabolic responses to unloading and reloading.

Management of immobilization and its complication for elderly.

Science.gov (United States)

Laksmi, Purwita W; Harimurti, Kuntjoro; Setiati, Siti; Soejono, Czeresna H; Aries, Wanarani; Roosheroe, Arya Govinda

2008-10-01

Increased life expectancy have an effect on the rising percentage of elderly population in Indonesia and health problem associated with the elderly, particularly immobilization. Immobilization may cause various complications, especially when it has been overlooked without any appropriate and proper medical care in keeping with the procedures. High incidence of immobilization in elderly and the life-threatening complication call for an agreement on management of immobilization and its complication. Management of immobilization needs interdisciplinary team-work cooperation, the patients and their family. The management may be commenced through a complete geriatric review, formulating functional goals and constructing therapeutic plan. Various medical conditions and external factors that may act as risk factors of immobilization as well as drugs intake that may exaggerate the immobilization should be evaluated and optimally managed. Any complication due to immobilization and other concomitant disease/condition should be recognized and managed comprehensively in order to reduce morbidity and mortality. Management of immobilization and its complications include pharmacological and non-pharmacological treatment, i.e. various mobility exercises, utilization of ambulatory device and supporting appliance for assisting patients in stand-up position, as well as the management of urinary voiding and defecation.

Technetium Immobilization Forms Literature Survey

Energy Technology Data Exchange (ETDEWEB)

Westsik, Joseph H.; Cantrell, Kirk J.; Serne, R. Jeffrey; Qafoku, Nikolla

2014-05-01

Of the many radionuclides and contaminants in the tank wastes stored at the Hanford site, technetium-99 (99Tc) is one of the most challenging to effectively immobilize in a waste form for ultimate disposal. Within the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the Tc will partition between both the high-level waste (HLW) and low-activity waste (LAW) fractions of the tank waste. The HLW fraction will be converted to a glass waste form in the HLW vitrification facility and the LAW fraction will be converted to another glass waste form in the LAW vitrification facility. In both vitrification facilities, the Tc is incorporated into the glass waste form but a significant fraction of the Tc volatilizes at the high glass-melting temperatures and is captured in the off-gas treatment systems at both facilities. The aqueous off-gas condensate solution containing the volatilized Tc is recycled and is added to the LAW glass melter feed. This recycle process is effective in increasing the loading of Tc in the LAW glass but it also disproportionally increases the sulfur and halides in the LAW melter feed which increases both the amount of LAW glass and either the duration of the LAW vitrification mission or the required supplemental LAW treatment capacity.

Technologies for destruction of long-lived radionuclides in high-level nuclear waste: Overview and requirements

International Nuclear Information System (INIS)

Arthur, E.D.

1993-01-01

This paper, and this topical session on Nuclear Waste Minimization, Management and Remediation, focuses on two nuclear systems, and their associated technologies, that have the potential to address concerns surrounding long-lived radionuclides in high-level waste. Both systems offer technology applicable to HLW from present light-water reactors (LWR). Additionally these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes, long-term fuel supplies, and enhanced safety. The first system is the Integral Fast Reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while simultaneously burning long-lived actinide waste. IFR applications include burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, Accelerator Transmutation of Waste (ATW), is based upon an accelerator-induced intense source of thermal neutrons and is aimed at destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

Immobilization of simulated radionuclide 133Cs{sup +} by fly ash-based geopolymer

Energy Technology Data Exchange (ETDEWEB)

Li, Qin; Sun, Zengqing; Tao, Dejing; Xu, Yan; Li, Peiming; Cui, Hao; Zhai, Jianping, E-mail: jpzhai@nju.edu.cn

2013-11-15

Highlights: • Fly ash-based geopolymer was used to immobilize 133Cs{sup +}, with cement as comparison. • Less Cs{sup +} was leached out from geopolymer in deionized water, acid and salt solutions. • Geopolymer showed more excellent acid resistance than cement blocks. • Geopolymer maintained superior mechanical strength to cement matrices. • Geopolymer showed good freeze–thaw and high-temperature performances. -- Abstract: The recent nuclear leak in Japan once again attracted people's attention to nuclear safety problems. Because of their poor thermal stability, those low-cost materials such as cement and asphalt cannot be used for the solidification of the radioactive wastes. In this work, the solidification behavior of 133Cs{sup +} by fly ash-based geopolymer was investigated. Leaching tests (carried out in deionized water, sulfuric acid and magnesium sulfate solutions) revealed that the geopolymer solidification had lower cumulative fraction leaching concentration (CFLC) of 133Cs{sup +} than that of cemented form. The thermal stability (high-temperature and freeze–thaw resistance) and acid-resistance of the geopolymer were also both better than that of cement. The geopolymer solidification block can acquire a compressive strength up to 30 MPa after 2 h calcination at 1000 °C. The morphology and mineral phases of the geopolymer and the geopolymer solidification block were characterized by SEM and XRD, and EDX analysis indicated that most of Cs associated with the amorphous geopolymer gel. These results gave encouragement for the idea that the fly ash-based geopolymer could be used as a low-cost and high-efficiency material for the immobilization of radioactive wastes.

Strength of interactions between immobilized dye molecules and sol-gel matrices.

Science.gov (United States)

Ismail, Fanya; Schoenleber, Monika; Mansour, Rolan; Bastani, Behnam; Fielden, Peter; Goddard, Nicholas J

2011-02-21

In this paper we present a new theory to re-examine the immobilization technique of dye doped sol-gel films, define the strength and types of possible bonds between the immobilized molecule and sol-gel glass, and show that the immobilized molecule is not free inside the pores as was previously thought. Immobilizing three different pH sensitive dyes with different size and functional groups inside the same sol-gel films revealed important information about the nature of the interaction between the doped molecule and the sol-gel matrix. The samples were characterized by means of ultraviolet-visible spectrophotometer (UV-VIS), thermal gravimetric analysis (TGA), mercury porosimetry (MP), nuclear magnetic resonance spectroscopy ((29)Si NMR) and field-emission environmental scanning electron microscopy (ESEM-FEG). It was found that the doped molecule itself has a great effect on the strength and types of the bonds. A number of factors were identified, such as number and types of the functional groups, overall charge, size, pK(a) and number of the silanol groups which surround the immobilized molecule. These results were confirmed by the successful immobilization of bromocresol green (BCG) after a completely polymerized sol-gel was made. The sol-gel consisted of 50% tetraethoxysilane (TEOS) and 50% methyltriethoxysilane (MTEOS) (w/w). Moreover, the effect of the immobilized molecule on the structure of the sol-gel was studied by means of a leaky waveguide (LW) mode for doped films made before and after polymerization of the sol-gel.

Triclosan-immobilized polyamide thin film composite membranes with enhanced biofouling resistance

Science.gov (United States)

Park, Sang-Hee; Hwang, Seon Oh; Kim, Taek-Seung; Cho, Arah; Kwon, Soon Jin; Kim, Kyoung Taek; Park, Hee-Deung; Lee, Jung-Hyun

2018-06-01

We report on a strategy to improve biofouling resistance of a polyamide (PA) thin-film composite (TFC) reverse osmosis (RO) membrane via chemically immobilizing triclosan (TC), known as a common organic biocide, on its surface. To facilitate covalent attachment of TC on the membrane surface, TC was functionalized with amine moiety to prepare aminopropyl TC. Then, the TC-immobilized TFC (TFC-TC) membranes were fabricated through a one-step amide formation reaction between amine groups of aminopropyl TC and acyl chloride groups present on the PA membrane surface, which was confirmed by high-resolution XPS. Strong stability of the immobilized TC was also confirmed by a hydraulic washing test. Although the TFC-TC membrane showed slightly reduced separation performance compared to the pristine control, it still maintained a satisfactory RO performance level. Importantly, the TFC-TC membrane exhibited excellent antibacterial activity against both gram negative (E. coli and P. aeruginosa) and gram positive (S. aureus) bacteria along with greatly enhanced resistance to biofilm formation. Our immobilization approach offers a robust and relatively benign strategy to control biofouling of functional surfaces, films and membranes.

Spent nuclear fuel project high-level information management plan

Energy Technology Data Exchange (ETDEWEB)

Main, G.C.

1996-09-13

This document presents the results of the Spent Nuclear Fuel Project (SNFP) Information Management Planning Project (IMPP), a short-term project that identified information management (IM) issues and opportunities within the SNFP and outlined a high-level plan to address them. This high-level plan for the SNMFP IM focuses on specific examples from within the SNFP. The plan`s recommendations can be characterized in several ways. Some recommendations address specific challenges that the SNFP faces. Others form the basis for making smooth transitions in several important IM areas. Still others identify areas where further study and planning are indicated. The team`s knowledge of developments in the IM industry and at the Hanford Site were crucial in deciding where to recommend that the SNFP act and where they should wait for Site plans to be made. Because of the fast pace of the SNFP and demands on SNFP staff, input and interaction were primarily between the IMPP team and members of the SNFP Information Management Steering Committee (IMSC). Key input to the IMPP came from a workshop where IMSC members and their delegates developed a set of draft IM principles. These principles, described in Section 2, became the foundation for the recommendations found in the transition plan outlined in Section 5. Availability of SNFP staff was limited, so project documents were used as a basis for much of the work. The team, realizing that the status of the project and the environment are continually changing, tried to keep abreast of major developments since those documents were generated. To the extent possible, the information contained in this document is current as of the end of fiscal year (FY) 1995. Programs and organizations on the Hanford Site as a whole are trying to maximize their return on IM investments. They are coordinating IM activities and trying to leverage existing capabilities. However, the SNFP cannot just rely on Sitewide activities to meet its IM requirements

Nuclear waste forms for actinides

Science.gov (United States)

Ewing, Rodney C.

1999-01-01

The disposition of actinides, most recently 239Pu from dismantled nuclear weapons, requires effective containment of waste generated by the nuclear fuel cycle. Because actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on risk assessments of geologic repositories. Thus, demonstrable, long-term chemical and mechanical durability are essential properties of waste forms for the immobilization of actinides. Mineralogic and geologic studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely materials science approach. The “mineralogic approach” is illustrated by a discussion of zircon as a phase for the immobilization of excess weapons plutonium. PMID:10097054

Retrievability of high-level nuclear waste from geologic repositories - Regulatory and rock mechanics/design considerations

International Nuclear Information System (INIS)

Tanious, N.S.; Nataraja, M.S.; Daemen, J.J.K.

1987-01-01

Retrievability of nuclear waste from high-level geologic repositories is one of the performance objectives identified in 10CFR60 (Code of Federal Regulations, 1985). 10CFR60.111 states that the geologic repository operations area shall be designed to preserve the option of waste retrieval. In designing the repository operations area, rock mechanics considerations play a major role especially in evaluating the feasibility of retrieval operations. This paper discusses generic considerations affecting retrievability as they relate to repository design, construction, and operation, with emphasis on regulatory and rock mechanics aspects

Immobilization of radioactive wastes in glasses and ceramics

International Nuclear Information System (INIS)

Zanotto, E.D.

1983-01-01

A large amount of radioactive liquid wastes arises from the reprocessing of spent nuclear fuels to recover uranium and plutonium. Immobilization of such wastes in solid form and disposal of the solidified wastes in safe places, to prevent contamination of the human environment, are topics of considerable interest for both the scientific community and the public in general. The great majority of materials candidate for the encapsulation of radioactive wastes are inorganic non-metalic, such as glasses, glass-ceramics, special cements, calcined ceramics and few more. Among these materials, certain glasses have received special attention, and are being studied for over twenty years. It is estimated that about US$2 billion have already been spent in these studies. The disposal (long term storage) of these solid wastes may be possible in deep geological formations, salt mines, the ocean bed, by evacuation to the outer space, etc. A brief review on the several options avaiable for encapsulation and disposal of high level radioactive liquid wastes is presented, along with the relative merits and disadvantages of the candidate materials for encapsulation. A few suggestions for the solution of the Brazilian problem are advanced. (Author) [pt

Methods of Disposing Of High-Level Radioactive Waste: A Review

International Nuclear Information System (INIS)

Abumurade, K.

2002-01-01

High level nuclear waste from both commercial reactors and defense industry presents a difficult problem to the scientific community as well as the public. The solutions to this problem is still debatable both technically and ethically. There are few methods proposed for disposing of high level waste. Each method has its own advantages and disadvantages. However, the very deep underground geologic repository is the best choice for disposing of high-level radioactive wastes. The cost benefit equation of nuclear power production and its waste is discussed. However, the public should be educated about this matter to minimize the gap between them and the nuclear power community including scientists industry, and governments. (Author) 15 refs., 4 tabs., 1 fig

Integrated development and testing plan for the plutonium immobilization project

International Nuclear Information System (INIS)

Kan, T.

1998-01-01

plutonium-containing ceramic forms within large canisters of high level waste (HLW) glass. Deployment of the immobilization capability should occur by 2006 and be completed within 10 years

IMMOBILIZATION OF FLAVIN ON HIGHLY POROUS POLYMERIC DISKS - 3 ROUTES TO A CATALYTICALLY ACTIVE MEMBRANE

NARCIS (Netherlands)

SCHOO, HFM; CHALLA, G; ROWATT, B; SHERRINGTON, DC

Disks obtained by polymerization of high internal phase emulsions (Polyhipe) had completely open pore structures and were used as a carrier material for the immobilisation of 10-ethyl-isoalloxazine ('flavin'). Three methods for immobilization are described: (1) direct modification of

Studies of Corrosion Resistant Materials Being Considered for High-Level Nuclear Waste Containment in Yucca Mountain Relevant Environments

International Nuclear Information System (INIS)

McCright, R.D.; Ilevbare, G.; Estill, J.; Rebak, R.

2001-01-01

Containment of spent nuclear fuel and vitrified forms of high level nuclear waste require use of materials that are highly corrosion resistant to all of the anticipated environmental scenarios that can occur in a geological repository. Ni-Cr-Mo Alloy 22 (UNS N60622) is proposed for the corrosion resistant outer barrier of a two-layer waste package container at the potential repository site at Yucca Mountain. A range of water compositions that may contact the outer barrier is under consideration, and a testing program is underway to characterize the forms of corrosion and to quantify the corrosion rates. Results from the testing support models for long term prediction of the performance of the container. Results obtained to date indicate a very low general corrosion rate for Alloy 22 and very high resistance to all forms of localized and environmentally assisted cracking in environments tested to date

Immobilization Patterns and Dynamics of Acetate-Utilizing Methanogens Immobilized in Sterile Granular Sludge in Upflow Anaerobic Sludge Blanket Reactors

Science.gov (United States)

Schmidt, Jens Ejbye; Ahring, Birgitte Kjær

1999-01-01

Sterile granular sludge was inoculated with either Methanosarcina mazeii S-6, Methanosaeta concilii GP-6, or both species in acetate-fed upflow anaerobic sludge blanket (UASB) reactors to investigate the immobilization patterns and dynamics of aceticlastic methanogens in granular sludge. After several months of reactor operation, the methanogens were immobilized, either separately or together. The fastest immobilization was observed in the reactor containing M. mazeii S-6. The highest effluent concentration of acetate was observed in the reactor with only M. mazeii S-6 immobilized, while the lowest effluent concentration of acetate was observed in the reactor where both types of methanogens were immobilized together. No changes were observed in the kinetic parameters (Ks and μmax) of immobilized M. concilii GP-6 or M. mazeii S-6 compared with suspended cultures, indicating that immobilization does not affect the growth kinetics of these methanogens. An enzyme-linked immunosorbent assay using polyclonal antibodies against either M. concilii GP-6 or M. mazeii S-6 showed significant variations in the two methanogenic populations in the different reactors. Polyclonal antibodies were further used to study the spatial distribution of the two methanogens. M. concilii GP-6 was immobilized only on existing support material without any specific pattern. M. mazeii S-6, however, showed a different immobilization pattern: large clumps were formed when the concentration of acetate was high, but where the acetate concentration was low this strain was immobilized on support material as single cells or small clumps. The data clearly show that the two aceticlastic methanogens immobilize differently in UASB systems, depending on the conditions found throughout the UASB reactor. PMID:10049862

The high level vibration test program

International Nuclear Information System (INIS)

Hofmayer, C.H.; Curreri, J.R.; Park, Y.J.; Kato, W.Y.; Kawakami, S.

1989-01-01

As part of cooperative agreements between the US and Japan, tests have been performed on the seismic vibration table at the Tadotsu Engineering Laboratory of Nuclear Power Engineering Test Center (NUPEC) in Japan. The objective of the test program was to use the NUPEC vibration table to drive large diameter nuclear power piping to substantial plastic strain with an earthquake excitation and to compare the results with state-of-the-art analysis of the problem. The test model was subjected to a maximum acceleration well beyond what nuclear power plants are designed to withstand. A modified earthquake excitation was applied and the excitation level was increased carefully to minimize the cumulative fatigue damage due to the intermediate level excitations. Since the piping was pressurized, and the high level earthquake excitation was repeated several times, it was possible to investigate the effects of ratchetting and fatigue as well. Elastic and inelastic seismic response behavior of the test model was measured in a number of test runs with an increasing excitation input level up to the limit of the vibration table. In the maximum input condition, large dynamic plastic strains were obtained in the piping. Crack initiation was detected following the second maximum excitation run. Crack growth was carefully monitored during the next two additional maximum excitation runs. The final test resulted in a maximum crack depth of approximately 94% of the wall thickness. The HLVT (high level vibration test) program has enhanced understanding of the behavior of piping systems under severe earthquake loading. As in other tests to failure of piping components, it has demonstrated significant seismic margin in nuclear power plant piping

Process arrangement options for Defense waste immobilization

International Nuclear Information System (INIS)

1980-02-01

Current plans are to immobilize the SRP high-level liquid wastes in a high integrity form. Borosilicate glass was selected in 1977 as the reference waste form and a mjaor effort is currently underway to develop the required technology. A large new facility, referred to as the Defense Waste Processing Facility (DWPF) is being designed to carry out this mission, with project authorization targeted for 1982 and plant startup in 1989. However, a number of other process arrangements or manufacturing strategies, including staging the major elements of the project or using existing SRP facilities for some functions, have been suggested in lieu of building the reference DWPF. This study assesses these various options and compares them on a technical and cost basis with the DWPF. Eleven different manufacturing options for SRP defense waste solidification were examined in detail. These cases are: (1) vitrification of acid waste at current generation rate; (2) vitrification of current rate acid waste and caustic sludge; (3 and 4) vitrification of the sludge portion of neutralized waste; (5) decontamination of salt cake and storage of concentrated cesium and strontium for later immobilization; (6) processing waste in a facility with lower capacity than the DWPF; (7) processing waste in a combination of existing and new facilities; (8) waste immobilization in H Canyon; (9) vitrification of both sludge and salt; (10) DWPF with onsite storage; (11) deferred authorization of DWPF

Technologies for immobilization and disposal of tritium

International Nuclear Information System (INIS)

Coppari, N.R.

1996-01-01

This study was done within a program one of whose objectives was to know the state of the technology development for tritium separation in the moderator circuit at HWR and to define the possible technologies to be applied to the Argentine nuclear power plants. Within this framework the strategies adopted by each country and the available technologies for a safe disposal of tritium, not only in its gaseous state tritium but also as tritiated water were analyzed. It is considered that if the selected separation method is such that the tritium is in its gaseous state, the hydride formation for long periods of immobilization should be studied. whereas if it were triated water immobilization should be studied to choose the technology between cementation and drying agents, in both cases the final disposal site will have to be selected. (author). 8 refs

The Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Dixon, R.S.

1984-12-01

The Canadian Nuclear Fuel Waste Management Program involves research into the storage and transportation of used nuclear fuel, immobilization of fuel waste, and deep geological disposal of the immobilized waste. The program is now in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the safety and environmental aspects of the deep underground disposal of immobilized fuel waste in plutonic rock. The objectives of the research for each component of the program and the progress made to the end of 1983 are described in this report

Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

Energy Technology Data Exchange (ETDEWEB)

Rashid, F.; Ghaffar, A. [Pakistan Inst. of Nuclear Science and Technology, Islamabad (Pakistan)

2011-07-01

The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

International Nuclear Information System (INIS)

Rashid, F.; Ghaffar, A.

2011-01-01

The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

Californium-252 neutron activation analysis of high-level processed nuclear tank waste

International Nuclear Information System (INIS)

Troyer, G.L.; Purcell, M.A.

2000-01-01

The basis for production assessment of the vitrification of Hanford nuclear fuel reprocessing wastes will be high-precision measurements of the elemental sodium content. However, the chemical analysis of both radioactive and nonradioactive components in nuclear waste can be challenged by high radiation dose rates. The dose rates compromise many analytical techniques as well as pose personnel dosimetry risks. In many cases, reduction of dose rates through dilution compromises the precision and sensitivity for certain key components. The use of neutron activation analysis (NAA) provides a method of analysis that avoids the need for dilutions or extensive sample preparation. These waste materials also contain trace quantities of fissionable isotopes, which, through neutron activation, can be estimated by delayed neutron counting of fissioned fragments

High photocatalytic activity of immobilized TiO{sub 2} nanorods on carbonized cotton fibers

Energy Technology Data Exchange (ETDEWEB)

Wang, Bin, E-mail: bwang23@cityu.edu.hk [Ability R and D Energy Research Center, School of Energy and Environment, City University of Hong Kong, Hong Kong (China); Karthikeyan, Rengasamy; Lu, Xiao-Ying [Ability R and D Energy Research Center, School of Energy and Environment, City University of Hong Kong, Hong Kong (China); Xuan, Jin [Ability R and D Energy Research Center, School of Energy and Environment, City University of Hong Kong, Hong Kong (China); State-Key Laboratory of Chemical Engineering, School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China); Leung, Michael K.H., E-mail: mkh.leung@cityu.edu.hk [Ability R and D Energy Research Center, School of Energy and Environment, City University of Hong Kong, Hong Kong (China)

2013-12-15

Highlights: • Hollow carbon fibers derived from natural cotton was successfully prepared by pyrolysis method. • TiO{sub 2} nanorods immobilized on carbon fibers by a facile hydrothermal method showed high photocatalytic activity. • The enhancement was due to the reduced band gap, improved dye adsorption capacity and effective electron–hole separation. -- Abstract: In this study, TiO{sub 2} nanorods were successfully immobilized on carbon fibers by a facile pyrolysis of natural cotton in nitrogen atmosphere followed by a one-pot hydrothermal method. Carbonized cotton fibers (CCFs) and TiO{sub 2}-CCFs composites were characterized using field-emission scanning electron microscope (FE-SEM), transmission electron microscopy (TEM), X-ray photoelectron spectroscopy (XPS), Raman spectroscopy, X-ray diffractometer (XRD), diffuse reflectance UV–vis spectroscopy (DRS) and photoluminescence (PL) spectroscopy. Results implied that the band gap narrowing of TiO{sub 2} was achieved after integration of CCFs. Dye adsorption isotherm indicated that the maximum dye adsorption capacity (q{sub m}) of CCFs-1000 (13.4 mg/g) was 2 times higher than that of cotton fibers and q{sub m} of TiO{sub 2}-CCFs-1000 (9.0 mg/g) was 6–7 times higher than that of TiO{sub 2} nanorods. Photocatalytic activity of TiO{sub 2} nanorods prepared with 3 mL Ti(OBu){sub 4} showed the highest photocatalytic activity. TiO{sub 2}-CCFs-1000 exhibited higher activity than TiO{sub 2} immobilized on CCFs-400, CCFs-600 and CCFs-800. Good photostability of TiO{sub 2}-CCFs-1000 was found for dye degradation under visible light irradiation. The enhancement of photocatalytic dye degradation was due to the high adsorptivity of dye molecules, enhanced light adsorption and effective separation of electron–hole pairs. This work provides a low-cost and sustainable approach to immobilize nanostructured TiO{sub 2} on carbon fibers for environmental remediation.

Immobilization of microorganisms. Part 1. Preparation of immobilized Lactobacillus bulgaricus

Energy Technology Data Exchange (ETDEWEB)

Lee, K H

1981-01-01

The immobilization of Lactobacillus bulgaricus on polyacrylamide and on alginate beads was investigated. The most active immobilized cells were obtained by entrapment in Ca alginate beads. These immobilized microbial cells, when introduced into 4.5% lactose solution and whey solution showed maximum relative activity of 28% for lactose and 18% for whey compared to free cells.

Salton Sea geothermal field as a natural analog for the near-field in a salt high-level nuclear waste repository

International Nuclear Information System (INIS)

Elders, W.A.; Moody, J.B.; Battelle Memorial Inst., Columbus, OH)

1984-01-01

The Salton Sea Geothermal Field (SSGF), on the delta of the Colorado River in southern California, is being studied as a natural analog for the near-field environment of proposed nuclear waste repositories in salt. A combination of mineralogical and geochemical methods is being employed to develop a three-dimenisonal picture of temperature, salinity, lithology, mineralogy, and chemistry of reactions between the reservoir rocks and the hot brines. Our aim is to obtain quantitative data on mineral stabilities and on mobilities of the naturally occurring radionuclides of concern in Commercial High-Level Waste (CHLW). These data will be used to validate the EQ3/6 geochemical code under development to model the salt near-field repository behavior. Maximum temperatures encountered in wells in the SSGF equal or exceed peak temperatures expected in a salt repository. Brines produced from these wells have major element chemistry similar to brines from candidate salt sites. Relative to the rocks, these brines are enriched in Na, Mn, Sr, Ra, and Po, depleted in Ba, Si, Mg, Ti, and Al, and strongly depleted in U and Th. However, the unaltered rocks contain only about 2 to 3 ppm of U and 4 to 12 ppm of Th, largely in detrital epidotes and zircons. Samples of hydrothermally altered rocks from a wide range of temperature and salinity show rather similar uniform low concentrations of these elements, even when authigenic illite, chlorite, ipidote and feldspar are present. These observations suggest that U and Th are relatively immobile in these hot brines. However, Ra, Po, Cs, and Sr are relatively mobile. Work is continuing to document naturally occurring radionuclide partitioning between SSGF minears and brine over a range of temperature, salinity, and lithology. 8 refs., 7 figs., 2 tabs

Nuclear systems of level measurement

International Nuclear Information System (INIS)

Lara, A.J.; Cabrera, M.J.

1992-01-01

In the industry there are processes in which is necessary to maintain the products level controlled which are handled for their transformation. The majority of such processes and by the operation conditions, they do not admit measure systems of level of invasive type then the application of nuclear techniques for level measurement results a big aid in these cases, since all the system installation is situated beyond frontiers of vessels that contain the product for measuring. In the Department of Nuclear Technology Applications of Mexican Petroleum Institute was developed a level measurement system by gamma rays transmission which operates in the Low Density Polyethylene plant of Petrochemical Complex Escolin at Poza Rica, Veracruz, Mexico. (Author)

Biological ramifications of the subseabed disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Gomez, L.S.; Hessler, R.R.; Jackson, D.W.; Marietta, M.G.; Smith, K.L. Jr.; Talbert, D.M.; Yayanos, A.A.

1980-01-01

The primary goal of the US Subseabed Disposal Program (SDP) is to assess the technical and environmental feasibility of disposing of high-level nuclear waste in deep-sea sediments. The subseabed biology program is charged with assessing possible ecosystem effects of radionuclides as well as possible health effects to man from radionuclides which may be released in the deep sea and transported to the ocean surface. Current biological investigations are attempting to determine benthic community structure; benthic community metabolism; the biology of deep-sea mobile scavengers; the faunal composition of midwater nekton; rates of microbial processes, and the radiation sensitivity of deep-sea organisms. Existing models of the dispersal of radionuclides in the deep sea have not considered many of the possible biological mechanisms which may influence the movement of radionuclides. Therefore, a multi-compartment foodweb model is being developed which considers both biological and physical influences on radionuclide transport. This model will allow parametric studies to be made of the impact on the ocean environment and on man of potential releases of radionuclides

Biological ramifications of the subseabed disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Gomez, L.S.; Hessler, R.R.; Jackson, D.W.; Marietta, M.G.; Smith, K.L. Jr.; Talbert, D.M.; Yayanos, A.A.

1980-05-01

The primary goal of the US Subseabed Disposal Program (SDP) is to assess the technical and environmental feasibility of disposing of high-level nuclear waste in deep-sea sediments. The subseabed biology program is charged with assessing possible ecosystem effects of radionuclides as well as possible health effects to man from radionuclides which may be released in the deep sea and transported to the ocean surface. Current biological investigations are attempting to determine benthic community structure; benthic community metabolism; the biology of deep-sea mobile scavengers; the faunal composition of midwater nekton; rates of microbial processes; and the radiation sensitivity of deep-sea organisms. Existing models of the dispersal of radionuclides in the deep sea have not considered many of the possible biological mechanisms which may influence the movement of radionuclides. Therefore, a multi-compartment foodweb model is being developed which considers both biological and physical influences on radionuclide transport. This model will allow parametric studies to be made of the impact on the ocean environment and on man of potential releases of radionuclides

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

International Nuclear Information System (INIS)

Hamilton, L.

2001-01-01

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

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

International Nuclear Information System (INIS)

Hamilton, L.

2001-01-01

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

Automated, High Temperature Furnace for Glovebox Operation

International Nuclear Information System (INIS)

Neikirk, K.

2001-01-01

The U.S. Department of Energy will immobilize excess plutonium in the proposed Plutonium Immobilization Plant (PIP) at the Savannah River Site (SRS) as part of a two track approach for the disposition of weapons usable plutonium. As such, the Department of Energy is funding a development and testing effort for the PIP. This effort is being performed jointly by Lawrence Livermore National Laboratory (LLNL), Westinghouse Savannah River Company (WSRC), Pacific Northwest National Laboratory (PNNL), and Argonne National Laboratory (ANL). The Plutonium Immobilization process involves the disposition of excess plutonium by incorporation into ceramic pucks. As part of the immobilization process, furnaces are needed for sintering the ceramic pucks. The furnace being developed for puck sintering is an automated, bottom loaded furnace with insulting package and resistance heating elements located within a nuclear glovebox. Other furnaces considered for the application include retort furnaces and pusher furnaces. This paper, in part, will discuss the furnace technologies considered and furnace technology selected to support reliable puck sintering in a glovebox environment. Due to the radiation levels and contamination associated with the plutonium material, the sintering process will be fully automated and contained within nuclear material gloveboxes. As such, the furnace currently under development incorporates water and air cooling to minimize heat load to the glovebox. This paper will describe the furnace equipment and systems needed to employ a fully automated puck sintering process within nuclear gloveboxes as part of the Plutonium Immobilization Plant

Enhanced nuclear level decay in hot dense plasmas

International Nuclear Information System (INIS)

Gosselin, G.; Morel, P.

2004-01-01

A model of nuclear level decay in a plasma environment is described. Nuclear excitation and decay by photon processes, nuclear excitation by electron capture, and decay by internal conversion are taken into account. The electrons in the plasma are described by a relativistic average atom model for the bound electrons and by a relativistic Thomas-Fermi-Dirac model for the free electrons. Nuclear decay of isomeric level may be enhanced through an intermediate level lying above the isomer. An enhanced nuclear decay rate may occur for temperatures far below the excitation energy of the transition to the intermediate level. In most cases, the enhancement factor may reach several decades

Evaluation of sulfur polymer cement as a waste form for the immobilization of low-level radioactive or mixed waste

International Nuclear Information System (INIS)

Mattus, C.H.; Mattus, A.J.

1994-03-01

Sulfur polymer cement (SPC), also called modified sulphur cements, is a relatively new material in the waste immobilization field, although it was developed in the late seventies by the Bureau of Mines. The physical and chemical properties of SPC are interesting (e.g., development of high mechanical strength in a short time and high resistance to many corrosive environments). Because of its very low permeability and porosity, SPC is especially impervious to water, which, in turn, has led to its consideration for immobilization of hazardous or radioactive waste. Because it is a thermosetting process, the waste is encapsulated by the sulfur matrix; therefore, very little interaction occurs between the waste species and the sulfur (as there can be when waste prevents the set of portland cement-based waste forms)

Nuclear power: levels of safety

International Nuclear Information System (INIS)

Lidsky, L.M.

1988-01-01

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the nuclear establishment itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired

Thirty Years of Social Science Research on High-Level Nuclear Waste: Achievements and Future Challenges

International Nuclear Information System (INIS)

Solomon, Barry D.; Andren, Mats; Strandberg, Urban

2010-09-01

Research on high-level nuclear waste management has focused on technical and scientific issues since the U.S. National Academy of Sciences first studied the problem in the mid 1950s and recommended long-term disposal in deep salt formations. In this review, we trace the development of the problem's definition and its associated research since socioeconomic, political and policy issues were first given consideration and nuclear waste management became recognized as more than a technical issue. Three time periods are identified. First, from the mid 1970s to early 1980s, initial research explored institutional dimensions of nuclear waste, including ethics. The second period began in the early 1980s with a concerted effort to solve the problem and site nuclear waste repositories, and ended in the mid 1990s with minimal progress in the U.S. and general stalemate in Asia and Europe (with the notable exception of Sweden). This phase accelerated research on risk perception and stigma of nuclear waste, and elevated a focus on public trust. Great attention was given to repository siting conflicts, while minimal attention was placed on ethics, equity, political systems, and public participation. The last period, since the mid 1990s, has been characterized by continuing political stalemate and increased attention to public participation, political systems and international solutions. Questions of ethics have been given renewed attention, while research on risk perceptions and siting conflicts continues. We frame these periods in a broader context of the shifting role of applied social scientists. The paper concludes with a general discussion of this research area and prospects for future research

Thirty Years of Social Science Research on High-Level Nuclear Waste: Achievements and Future Challenges

Energy Technology Data Exchange (ETDEWEB)

Solomon, Barry D. (Dept. of Social Sciences, Michigan Technological Univ., Houghton (United States)), e-mail: bdsolomo@mtu.edu; Andren, Mats; Strandberg, Urban (Center for Public Sector Research, Univ. of Goeteborg, Goeteborg (Sweden))

2010-09-15