Health and environmental risk-related impacts of actinide burning on high-level waste disposal

International Nuclear Information System (INIS)

Forsberg, C.W.

1992-05-01

The potential health and environmental risk-related impacts of actinide burning for high-level waste disposal were evaluated. Actinide burning, also called waste partitioning-transmutation, is an advanced method for radioactive waste management based on the idea of destroying the most toxic components in the waste. It consists of two steps: (1) selective removal of the most toxic radionuclides from high-level/spent fuel waste and (2) conversion of those radionuclides into less toxic radioactive materials and/or stable elements. Risk, as used in this report, is defined as the probability of a failure times its consequence. Actinide burning has two potential health and environmental impacts on waste management. Risks and the magnitude of high-consequence repository failure scenarios are decreased by inventory reduction of the long-term radioactivity in the repository. (What does not exist cannot create risk or uncertainty.) Risk may also be reduced by the changes in the waste characteristics, resulting from selection of waste forms after processing, that are superior to spent fuel and which lower the potential of transport of radionuclides from waste form to accessible environment. There are no negative health or environmental impacts to the repository from actinide burning; however, there may be such impacts elsewhere in the fuel cycle

Literature review of intrinsic actinide colloids related to spent fuel waste package release rates

Energy Technology Data Exchange (ETDEWEB)

Zhao, P.; Steward, S.A.

1997-01-01

Existence of actinide colloids provides an important mechanism in the migration of radionuclides and will be important in performance of a geologic repository for high-level nuclear waste. Actinide colloids have been formed during long-term unsaturated dissolution of spent fuel by groundwater. This article summarizes a literature search of actinide colloids. This report emphasizes the formation of intrinsic actinide colloids, because they would have the opportunity to form soon after groundwater contact with the spent fuel and before actinide-bearing groundwater reaches the surrounding geologic formations.

Literature review of intrinsic actinide colloids related to spent fuel waste package release rates

International Nuclear Information System (INIS)

Zhao, P.; Steward, S.A.

1997-01-01

Existence of actinide colloids provides an important mechanism in the migration of radionuclides and will be important in performance of a geologic repository for high-level nuclear waste. Actinide colloids have been formed during long-term unsaturated dissolution of spent fuel by groundwater. This article summarizes a literature search of actinide colloids. This report emphasizes the formation of intrinsic actinide colloids, because they would have the opportunity to form soon after groundwater contact with the spent fuel and before actinide-bearing groundwater reaches the surrounding geologic formations

Actinide recycle in LMFBRs as a waste management alternative

International Nuclear Information System (INIS)

Beaman, S.L.

1979-01-01

A strategy of actinide burnup in fast reactor systems has been investigated as an approach for reducing the long term hazards and storage requirements of the actinide waste elements and their decay daughters. The actinide recycle studies also included plutonium burnup studies in the event that plutonium is no longer required as a fuel. Particular emphasis was placed upon the timing of the recycle program, the requirements for separability of the waste materials, and the impact of the actinides on the reactor operations and performance. It is concluded that actinide recycle and plutonium burnout are attractive alternative waste management concepts. 25 refs., 14 figs., 34 tabs

Radiochemical separation of actinides for their determination in environmental samples and waste products

Energy Technology Data Exchange (ETDEWEB)

Gleisberg, B [Nuclear Engineering and Analytics Rossendorf, Inc. (VKTA), Dresden (Germany)

1997-03-01

The determination of low level activities of actinides in environmental samples and waste products makes high demands on radiochemical separation methods. Artificial and natural actinides were analyzed in samples form the surrounding areas of NPP and of uranium mines, incorporation samples, solutions containing radioactive fuel, solutions and solids resutling from the process, and in wastes. The activities are measured by {alpha}-spectrometry and {gamma}-spectrometry. (DG)

Special actinide nuclides: Fuel or waste?

International Nuclear Information System (INIS)

Srinivasan, M.; Rao, K.S.; Dingankar, M.V.

1989-01-01

The special actinide nuclides such as Np, Cm, etc. which are produced as byproducts during the operation of fission reactors are presently looked upon as 'nuclear waste' and are proposed to be disposed of as part of high level waste in deep geological repositories. The potential hazard posed to future generations over periods of thousands of years by these long lived nuclides has been a persistent source of concern to critics of nuclear power. However, the authors have recently shown that each and every one of the special actinide nuclides is a better nuclear fuel than the isotopes of plutonium. This finding suggests that one does not have to resort to exotic neutron sources for transmuting/incinerating them as proposed by some researchers. Recovery of the special actinide elements from the waste stream and recycling them back into conventional fission reactors would eliminate one of the stigmas attached to nuclear energy

Synthesis of nuclear waste monazites, ideal actinide hosts for geologic disposal

International Nuclear Information System (INIS)

McCarthy, G.J.; White, W.B.; Pfoertsch, D.E.

1978-01-01

Monazite, an orthophosphate mineral of the lanthanides (Ln) and the actinides (An) U and Th, is a model for an ideal synthetic mineral waste form for geologic disposal of long-lived nuclear waste actinides. Natural monazites are known to have survived many of the conditions that might be inflicted on a nuclear waste repository by geological disruptions. High Th and U monazites with compositions typical of nuclear wastes have been synthesized with a routine calcination-pelletization-crystallization procedure. Charge balance for the Th 4+ → Ln 3+ substitution can be provided by either an equimolar Ca 2+ → Ln 3+ or Si 4+ → P 5+ substitution. For U 4+ → Ln 3+ , only the Ca 2+ → Ln 3+ substitution resulted in a phase-pure monazite. Unit cell parameter data were obtained for each nuclear waste monazite phase

Review of the treatment of actinides-bearing radioactive wastes

International Nuclear Information System (INIS)

Krause, H.

1983-01-01

Actinides bearing wastes are produced above all in the course of irradiated nuclear fuel reprocessing and during fabrication of mixed oxide fuel elements. Particular attention in research and development work must be paid to this type of waste, mainly on account of its longevity. In practical application, the specific character of the actinides bearing wastes has been largely recognized. Nevertheless, definitions and methods of treatment generally accepted worldwide are still missing today. This has no bearing as yet on present day treatment of radioactive wastes. But by the time of application of the breeder technology at the latest a special treatment concept should be available which complies with the high actinide contents and short precooling periods of the wastes

Process analytical chemistry applied to actinide waste streams

International Nuclear Information System (INIS)

Day, R.S.

1994-01-01

The Department of Energy is being called upon to clean up it's legacy of waste from the nuclear complex generated during the cold war period. Los Alamos National Laboratory is actively involved in waste minimization and waste stream polishing activities associated with this clean up. The Advanced Testing Line for Actinide Separations (ATLAS) at Los Alamos serves as a developmental test bed for integrating flow sheet development of nitric acid waste streams with process analytical chemistry and process control techniques. The wastes require processing in glove boxes because of the radioactive components, thus adding to the difficulties of making analytical measurements. Process analytical chemistry methods provide real-time chemical analysis in support of existing waste stream operations and enhances the development of new waste stream polishing initiatives. The instrumentation and methods being developed on ATLAS are designed to supply near-real time analyses on virtually all of the chemical parameters found in nitric acid processing of actinide waste. These measurements supply information on important processing parameters including actinide oxidation states, free acid concentration, interfering anions and metal impurities

Disposition of actinides released from high-level waste glass

International Nuclear Information System (INIS)

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

1994-01-01

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

Technical requirements for the actinide source-term waste test program

Energy Technology Data Exchange (ETDEWEB)

Phillips, M.L.F.; Molecke, M.A.

1993-10-01

This document defines the technical requirements for a test program designed to measure time-dependent concentrations of actinide elements from contact-handled transuranic (CH TRU) waste immersed in brines similar to those found in the underground workings of the Waste Isolation Pilot Plant (WIPP). This test program wig determine the influences of TRU waste constituents on the concentrations of dissolved and suspended actinides relevant to the performance of the WIPP. These influences (which include pH, Eh, complexing agents, sorbent phases, and colloidal particles) can affect solubilities and colloidal mobilization of actinides. The test concept involves fully inundating several TRU waste types with simulated WIPP brines in sealed containers and monitoring the concentrations of actinide species in the leachate as a function of time. The results from this program will be used to test numeric models of actinide concentrations derived from laboratory studies. The model is required for WIPP performance assessment with respect to the Environmental Protection Agency`s 40 CFR Part 191B.

Technical requirements for the actinide source-term waste test program

International Nuclear Information System (INIS)

Phillips, M.L.F.; Molecke, M.A.

1993-10-01

This document defines the technical requirements for a test program designed to measure time-dependent concentrations of actinide elements from contact-handled transuranic (CH TRU) waste immersed in brines similar to those found in the underground workings of the Waste Isolation Pilot Plant (WIPP). This test program wig determine the influences of TRU waste constituents on the concentrations of dissolved and suspended actinides relevant to the performance of the WIPP. These influences (which include pH, Eh, complexing agents, sorbent phases, and colloidal particles) can affect solubilities and colloidal mobilization of actinides. The test concept involves fully inundating several TRU waste types with simulated WIPP brines in sealed containers and monitoring the concentrations of actinide species in the leachate as a function of time. The results from this program will be used to test numeric models of actinide concentrations derived from laboratory studies. The model is required for WIPP performance assessment with respect to the Environmental Protection Agency's 40 CFR Part 191B

Stainless steel-zirconium alloy waste forms

International Nuclear Information System (INIS)

McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.

1996-01-01

An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ''noble'' nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation

Removal of actinides from selected nuclear fuel reprocessing wastes

International Nuclear Information System (INIS)

Navratil, J.D.; Thompson, G.H.

1979-01-01

The US Department of Energy awarded Oak Ridge National Laboratory a program to develop a cost-risk-benefit analysis of partitioning long-lived nuclides from waste and transmuting them to shorter lived or stable nuclides. Two subtasks of this program were investigated at Rocky Flats. In the first subtask, methods for solubilizing actinides in incinerator ash were tested. Two methods appear to be preferable: reaction with ceric ion in nitric acid or carbonate-nitrate fusion. The ceric-nitric acid system solubilizes 95% of the actinides in ash; this can be increased by 2 to 4% by pretreating ash with sodium hydroxide to solubilize silica. The carbonate-nitrate fusion method solubilizes greater than or equal to 98% of the actinides, but requires sodium hydroxide pretreatment. Two additional disadvantages are that it is a high-temperature process, and that it generates a lot of salt waste. The second subtask comprises removing actinides from salt wastes likely to be produced during reactor fuel fabrication and reprocessing. A preliminary feasibility study of solvent extraction methods has been completed. The use of a two-step solvent extraction system - tributyl phosphate (TBP) followed by extraction with a bidentate organophosphorous extractant (DHDECMP) - appears to be the most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and > 99.99% of the uranium. The second step using DHDECMP would remove > 99.91% of the americium and the remaining plutonium (> 99.98%) and other actinides from the acidified salt waste. 8 figures, 11 tables

Separation of actinides and lanthanides from acidic nuclear wastes by supported liquid membranes

International Nuclear Information System (INIS)

Danesi, P.R.; Chiarizia, R.; Rickert, P.; Horwitz, E.P.

1985-01-01

Supported liquid membranes, SLM, consisting of a solution of 0.25 M octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and 0.75 M tributylphosphate (TBP) in decalin absorbed on thin microporous polypropylene supports, have been studied for their ability to perform selective separations and concentrations of actinide and lanthanide ions from synthetic acidic nuclear wastes. The permeability coefficients of selected actinides (Am, Pu, U, Np) and of some of the other major components of the wastes have been measured using SLMs in flat-sheet and hollow-fiber configurations. The results have shown that with the thin (25 μm) flat-sheet SLMs, using Celgard 2500 as support, the membrane permeation process is mainly controlled by the rate of diffusion through the aqueous boundary layers. With the thicker (430 μm) hollow-fiber SLMs, using Accurel hollow-fibers as support, the membrane permeation process is controlled by the rate of diffusion through both the SLM and the aqueous boundary layers. Hollow-fibers SLMs exhibited lower permeability coefficients and longer life-times. The experiments have shown that the actinides can be very efficiently removed from the synthetic waste solutions to the point that the resulting solution could be considered a non-transuranic waste (less than 100 mCi/g of disposed form). The work has demonstrated that actinide removal from synthetic waste solutions is a feasible chemical process at the laboratory scale level

MICROBIAL TRANSFORMATIONS OF TRU AND MIXED WASTES: ACTINIDE SPECIATION AND WASTE VOLUME REDUCTION.

Energy Technology Data Exchange (ETDEWEB)

FRANCIS, A.J.; DODGE, C.J.

2006-11-16

The overall goals of this research project are to determine the mechanism of microbial dissolution and stabilization of actinides in Department of Energy's (DOE) TRU wastes, contaminated sludges, soils, and sediments. This includes (1) investigations on the fundamental aspects of microbially catalyzed radionuclide and metal transformations (oxidation/reduction reactions, dissolution, precipitation, chelation); (2) understanding of the microbiological processes that control speciation and alter the chemical forms of complex inorganic/organic contaminant mixtures; and (3) development of new and improved microbially catalyzed processes resulting in immobilization of metals and radionuclides in the waste with concomitant waste volume reduction.

MICROBIAL TRANSFORMATIONS OF TRU AND MIXED WASTES: ACTINIDE SPECIATION AND WASTE VOLUME REDUCTION

Energy Technology Data Exchange (ETDEWEB)

Francis, A.J.; Dodge, C.J.

2006-06-01

The overall goals of this research project are to determine the mechanism of microbial dissolution and stabilization of actinides in Department of Energy’s (DOE) TRU wastes, contaminated sludges, soils, and sediments. This includes (i) investigations on the fundamental aspects of microbially catalyzed radionuclide and metal transformations (oxidation/reduction reactions, dissolution, precipitation, chelation); (ii) understanding of the microbiological processes that control speciation and alter the chemical forms of complex inorganic/organic contaminant mixtures; and (iii) development of new and improved microbially catalyzed processes resulting in immobilization of metals and radionuclides in the waste with concomitant waste volume reduction.

MICROBIAL TRANSFORMATIONS OF TRU AND MIXED WASTES: ACTINIDE SPECIATION AND WASTE VOLUME REDUCTION

Energy Technology Data Exchange (ETDEWEB)

Francis, A.J.; Dodge, C.J.

2006-06-01

The overall goals of this research project are to determine the mechanism of microbial dissolution and stabilization of actinides in Department of Energy's (DOE) TRU wastes, contaminated sludges, soils, and sediments. This includes (1) investigations on the fundamental aspects of microbially catalyzed radionuclide and metal transformations (oxidation/reduction reactions, dissolution, precipitation, chelation); (2) understanding of the microbiological processes that control speciation and alter the chemical forms of complex inorganic/organic contaminant mixtures; and (3) development of new and improved microbially catalyzed processes resulting in immobilization of metals and radionuclides in the waste with concomitant waste volume reduction.

Feasibility studies of actinide recycle in LMFBRs as a waste management alternative

International Nuclear Information System (INIS)

Beaman, S.L.; Aitken, E.A.

1976-01-01

A strategy of actinide burnup in LMFBRs is being investigated as a waste management alternative to long term storage of high level nuclear waste. This strategy is being evaluated because many of the actinides in the waste from spent-fuel reprocessing have half-lives of thousands of years and an alternative to geological storage may be desired. From a radiological viewpoint, the actinides and their daughters dominate the waste hazard for decay times beyond about 400 years. Actinide burnup in LMFBRs may be an attractive alternative to geological storage because the actinides can be effectively transmuted to fission products which have significantly shorter half-lives. Actinide burnup in LMFBRs rather than LWRs is preferred because the ratio of fission reaction rate to capture reaction rate for the actinides is higher in an LMFBR, and an LMFBR is not so sensitive to the addition of the actinide isotopes. An actinide target assembly recycle scheme is evaluated to determine the effects of the actinides on the LMFBR performance, including local power peaking, breeding ratio, and fissile material requirements. Several schemes are evaluated to identify any major problems associated with reprocessing and fabrication of recycle actinide-containing assemblies. The overall efficiency of actinide burnout in LMFBRs is evaluated, and equilibrium cycle conditions are determined. It is concluded that actinide recycle in LMFBRs offers an attractive alternative to long term storage of the actinides, and does not significantly affect the performance of the host LMFBR. Assuming a 0.1 percent or less actinide loss during reprocessing, a 0.1 percent loss of less during fabrication, and proper recycle schemes, virtually all of the actinides produced by a fission reactor economy could be transmuted in fast reactors

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Materials Characterization Center. Second workshop on irradiation effects in nuclear waste forms. Summary report

International Nuclear Information System (INIS)

Weber, W.J.; Turcotte, R.P.

1982-01-01

The purpose of this second workshop on irradiations effects was to continue the discussions initiated at the first workshop and to obtain guidance for the Materials Characterization Center in developing test methods. The following major conclusions were reached: Ion or neutron irradiations are not substitutes for the actinide-doping technique, as described by the MCC-6 Method for Preparation and Characterization of Actinide-Doped Waste Forms, in the final evaluation of any waste form with respect to the radiation effects from actinide decay. Ion or neutron irradiations may be useful for screening tests or more fundamental studies. The use of these simulation techniques as screening tests for actinide decay requires that a correlation between ion or neutron irradiations and actinide decay be established. Such a correlation has not yet been established and experimental programs in this area are highly recommended. There is a need for more fundamental studies on dose-rate effects, temperature dependence, and the nature and importance of alpha-particle effects relative to the recoil nucleus in actinide decay. There are insufficient data presently available to evaluate the potential for damage from ionizing radiation in nuclear waste forms. No additional test methods were recommended for using ion or neutron irradiations to simulate actinide decay or for testing ionization damage in nuclear waste forms. It was recognized that additional test methods may be required and developed as more data become available. An American Society for Testing and Materials (ASTM) Task Group on the Simulation of Radiation Effects in Nuclear Waste Forms (E 10.08.03) was organized to act as a continuing vehicle for discussions and development of procedures, particularly with regard to ion irradiations

Transmutation of LWR waste actinides in thermal reactors

International Nuclear Information System (INIS)

Gorrell, T.C.

1979-01-01

Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles

Sequestering agents for the removal of actinides from waste streams

Energy Technology Data Exchange (ETDEWEB)

Raymond, K.N.; White, D.J.; Xu, Jide; Mohs, T.R. [Univ. of California, Berkeley, CA (United States)

1997-10-01

The goal of this project is to take a biomimetic approach toward developing new separation technologies for the removal of radioactive elements from contaminated DOE sites. To achieve this objective, the authors are investigating the fundamental chemistry of naturally occurring, highly specific metal ion sequestering agents and developing them into liquid/liquid and solid supported actinide extraction agents. Nature produces sideophores (e.g., Enterobactin and Desferrioxamine B) to selectivity sequester Lewis acidic metal ions, in particular Fe(III), from its surroundings. These chelating agents typically use multiple catechols or hydroxamic acids to form polydentate ligands that chelate the metal ion forming very stable complexes. The authors are investigating and developing analogous molecules into selective chelators targeting actinide(IV) ions, which display similar properties to Fe(III). By taking advantage of differences in charge, preferred coordination number, and pH stability range, the transition from nature to actinide sequestering agents has been applied to the development of new and highly selective actinide extraction technologies. Additionally, the authors have shown that these chelating ligands are versatile ligands for chelating U(VI). In particular, they have been studying their coordination chemistry and fundamental interactions with the uranyl ion [UO{sub 2}]{sup 2+}, the dominant form of uranium found in aqueous media. With an understanding of this chemistry, and results obtained from in vivo uranium sequestration studies, it should be possible to apply these actinide(IV) extraction technologies to the development of new extraction agents for the removal of uranium from waste streams.

Role of actinide behavior in waste management

International Nuclear Information System (INIS)

Bartlett, J.W.

1976-01-01

For purposes of assessing the safety of repositories of radioactive wastes placed in geologic isolation, actinide behavior in the environment has been interpreted in terms of five steps of prediction: analysis of repository stability; geosphere transport; the geosphere-biosphere interface; biosphere transport; and biosphere consequences. Each step in the analysis requires models of nuclide behavior and data on the physical and chemical properties of the radioactivity. The scope of information required in order to make reliable safety assessments has been outlined. All steps in the assessment process are coupled; reliable models and data are therefore needed for each step. The prediction phase of safety assessment is also coupled to activities concerned with waste treatment, selection of the final form of the waste, and selection of repository sites and designs. Results of the predictions can impact these activities

Phoenix type concepts for transmutation of LWR waste minor actinides

International Nuclear Information System (INIS)

Segev, M.

1994-01-01

A number of variations on the original Phoenix theme were studied. The basic rationale of the Phoenix incinerator is making oxide fuel of the LWR waste minor actinides, loading it in an FFTF-like subcritical core, then bombarding the core with the high current beam accelerated protons to generate considerable energy through spallation and fission reactions. As originally assessed, if the machine is fed with 1600 MeV protons in a 102 mA current, then 8 core modules are driven to transmute the yearly minor actinides waste of 75 1000 MW LWRs into Pu 238 and fission products; in a 2 years cycle the energy extracted is 100000 MW d/T. This performance cannot be substantiated in a rigorous analysis. A calculational consistent methodology, based on a combined execution of the Hermes, NCNP, and Korigen codes, shows, nonetheless that changes in the original Phoenix parameters can upgrade its performance.The original Phoenix contains 26 tons minor actinides in 8 core modules; 1.15 m 3 module is shaped for 40% neutron leakage; with a beam of 102 mA the 8 modules are driven to 100000 MW/T in 10.5 years, burning out the yearly minor actinide waste of 15 LWRs; the operation must be assisted by grid electricity. If the 1.15 m 3 module is shaped to allow only 28% leakage, then a beam of 102 mA will drive the 8 modules to 100000 MW/T in 3.5 years, burning out the yearly minor actinides waste of 45 LWRs. Some net grid electricity will be generated. If 25 tons minor actinides are loaded into 5 modules, each 1.72 m 3 in volume and of 24% leakage, then a 97 mA beam will drive the module to 100000 MW/T in 2.5 years, burning out the yearly minor actinides waste of 70 LWRs. A considerable amount of net grid electricity will be generated. If the lattice is made of metal fuel, and 26 tons minor actinides are loaded into 32 small modules, 0.17 m 3 each, then a 102 mA beam will drive the modules to 100000 MW/T in 2 years, burning out the yearly minor actinides waste of 72 LWRs. A considerable

Methods for separating actinides from reprocessing and refabrication plant wastes

International Nuclear Information System (INIS)

Tedder, D.W.; Finney, B.C.; Blomeke, J.O.

1979-01-01

Chemical processing flowsheets have been developed to partition actinides from all actinide-bearing LWR fuel reprocessing and refabrication plant wastes. These wastes include high-activity-level liquids, scrap recovery liquors, HEPA filters and incinerator ashes, and chemical salt wastes such as sodium carbonate scrub solutions, detergent cleanup streams, and alkaline off-gas scrubber liquors. The separations processes that were adopted for this study are based on solvent extraction, cation exchange chromatography, and leaching with Ce 4+ -HNO 3 solution

Minor actinide transmutation - a waste management option

International Nuclear Information System (INIS)

Koch, L.

1986-01-01

The incentive to recycle minor actinides results from the reduction of the long-term α-radiological risk rather than from a better utilization of the uranium resources. Nevertheless, the gain in generated electricity by minor actinide transmutation in a fast breeder reactor can compensate for the costs of their recovery and make-up into fuel elements. Different recycling options of minor actinides are discussed: transmutation in liquid metal fast breeder reactors (LMFBRs) is possible as long as plutonium is not recycled in light water reactors (LWRs). In this case a minor actinide burner with fuel of different composition has to be introduced. The development of appropriate minor actinide fuels and their properties are described. The irradiation experiments underway or planned are summarized. A review of minor actinide partitioning from the PUREX waste stream is given. From the present constraints of LMFBR technology a reduction of the long-term α-radiological risk by a factor of 200 is deduced relative to that from the direct storage of spent LWR fuel. Though the present accumulation of minor actinides is low, nuclear transmutation may be needed when nuclear energy production has grown. (orig.)

Evaluation of thorium based nuclear fuel. Actinide waste

International Nuclear Information System (INIS)

Wichers, V.A.

1995-06-01

Use of thorium based fuel has recently been proposed as a possible way to reduce the amount of actinide waste from nuclear power. To examine this possibility, burnup calculations were done of five once-through Thorium Heavy Water Reactor (THWR) systems, and three THWR systems with uranium recycle. The natural uranium once-through system was adopted as reference. The studied THWR fuel systems differed in the choice of fissile makeup fuel and exit burnup. The HWR was chosen because of its good neutron economy. Actinide waste production (in mass per GW e a) and radiotoxicity (in ALI per GW e a) for storage times up to 10 6 a were calculated for each system. The study shows that the THWR system with uranium recycle and High Enriched Uranium (U-235) makeup fuel performed best, producing both the lowest amount of plutonium and actinide waste with the lowest radiotoxicity. Relative to the natural uranium in HWR once-through system, radiotoxicity is reduced by a factor varying between 2 and 50 for the full range of storage times up to 10 6 a. (orig.)

Preliminary design and neutronic analysis of a laser fusion driven actinide waste burning hybrid reactor

International Nuclear Information System (INIS)

Berwald, D.H.; Duderstadt, J.J.

1979-01-01

The laser fusion driven actinide waste burner (LDAB) system investigated uses partitioned fission power reactor generated actinide wastes dissolved in a molten tin alloy as feed material (or fuel). A novel fuel processing concept based on the high-temperature precipitation of ''actinide--nitrides'' from a liquid tin solution is proposed. This concept will allow for fission product removal to be performed entirely within the device at high burnup. No attempt has been made to optimize this system, but potential performance is impressive. The equilibrium LDAB design consumes 7.6 MT/y of actinide waste. This corresponds to the waste output from 136 light water reactors [1000 MW (electric)]. The mean life of an actinide atom in the LDAB is only 4.5 y; and actinides, once charged to the LDAB, might be reprocessed fewer times during irradiation than in previously proposed systems

Significance of actinide chemistry for the long-term safety of waste disposal

International Nuclear Information System (INIS)

Kim, Jae Il

2006-01-01

A geochemical approach to the long-term safety of waste disposal is discussed in connection with the significance of actinides, which shall deliver the major radioactivity inventory subsequent to the relatively short-term decay of fission products. Every power reactor generates transuranic (TRU) elements: plutonium and minor actinides (Np, Am, Cm), which consist chiefly of long-lived nuclides emitting alpha radiation. The amount of TRU actinides generated in a fuel life period is found to be relatively small (about 1 wt% or less in spent fuel) but their radioactivity persists many hundred thousands years. Geological confinement of waste containing TRU actinides demands, as a result, fundamental knowledge on the geochemical behavior of actinides in the repository environment for a long period of time. Appraisal of the scientific progress in this subject area is the main objective of the present paper. Following the introductory discussion on natural radioactivities, the nuclear fuel cycle is briefly brought up with reference to actinide generation and waste disposal. As the long-term disposal safety concerns inevitably with actinides, the significance of the aquatic actinide chemistry is summarized in two parts: the fundamental properties relevant to their aquatic behavior and the geochemical reactions in nanoscopic scale. The constrained space of writing allows discussion on some examples only, for which topics of the primary concern are selected, e.g. apparent solubility and colloid generation, colloid-facilitated migration, notable speciation of such processes, etc. Discussion is summed up to end with how to make a geochemical approach available for the long-term disposal safety of nuclear waste or for the Performance Assessment (PA) as known generally

Sol-gel technology applied to alternative high-level waste forms development

International Nuclear Information System (INIS)

Angelini, P.; Stinton, D.P.; Vavruska, J.S.; Caputo, A.J.; Lackey, W.J.

1981-01-01

Sol-gel technology appears applicable to waste solidification. It is attractive for remote operation, and a variety of waste compositions and forms can be produced. Spheres and pellets of gel-derived Synroc waste forms were produced. Spheres of the Synroc-B type were coated with pyrolytic carbon and silicon carbide. Partitioning of actinides in Synroc-B was experimentally determined

Process for denitrating waste solutions containing nitric acid actinides simultaneously separating the actinides

International Nuclear Information System (INIS)

Gompper, K.

1984-01-01

The invention should reduce the acid and nitrate content of waste solutions containing nitric acid as much as possible, should reduce the total salt content of the waste solution, remove the actinides contained in it by precipitation and reduce the α radio-activity in the remaining solution, without having to worry about strong reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig.) [de

ANSTO's waste forms for the 31. century

International Nuclear Information System (INIS)

Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

2004-01-01

ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and 99 Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

ANSTO's waste forms for the 31. century

Energy Technology Data Exchange (ETDEWEB)

Vance, E R; Begg, B D; Day, R A; Moricca, S; Perera, D S; Stewart, M W. A.; Carter, M L; McGlinn, P J; Smith, K L; Walls, P A; Robina, M La

2004-07-01

ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

Nuclear transmutation of actinides other than fuel as a radioactive waste management scheme

International Nuclear Information System (INIS)

Cecille, L.; Hage, W.; Hettinger, H.; Mannone, F.; Mousty, F.; Schmidt, E.; Sola, A.; Huber, B.; Koch, L.

1977-01-01

The bulk of fission products in the high-level waste (HLW) decays to innocuous hazard levels within about 600 years. Actinide waste and a few fission products however represent a potential risk up to some hundreds of thousand of years. An alternative to the disposal of the whole HLW in geological formations is its fractionation, a nuclear transmutation of long-lived isotopes in fission reactors and a geological disposal of the other components. This solution would decrease the potential long-term risks of the geological waste disposal and would also accomodate to the demand of public opinion. The results of studies related to this management scheme are outlined with special reference to areas, where additional effort is required for realistic cost/benefit evaluations. Reactor physics calculations demonstrated the feasibility of actinide incineration in thermal and fast reactors. Obtained transmutation rates are sufficiently high to garantee acceptably small actinide inventories in the reactor in the case of self-generated actinide recycling. It appears that fast breeders could be used as transmutation devices without major additional reactor devlopment work. The thermal power rating of actinide fuel elements and the contribution of actinides and of minor amounts of lanthanide impurities to the neutron economy of the reactor has been evaluated. Sensitivity studies indicated that the results are dependent on the reactor operation mode and on the accuracy of the nuclear data. These calculations permitted the identification of isotopes for which cross section masurements and improved theoretical methods are required. The chemical separation of actinides from the HLW with the envisaged decontamination factors is being studied by solvent extraction and precipitation techniques using waste simulates and samples of high activity waste from European reprocessing plants. Up to now, the obtained results do not yet allow a definitive judgement on the feasibility of actinides

Recovery of actinides from TBP-Na2Co3 scrub-waste solutions: the ARALEX process

International Nuclear Information System (INIS)

Horwitz, E.P.; Bloomquist, C.A.A.; Mason, G.W.; Leonard, R.A.; Ziegler, A.A.

1979-08-01

A flowsheet for the recovery of actinides from TBP-Na 2 CO 3 scrub-waste solutions has been developed, based on batch extraction data, and tested, using laboratory-scale countercurrent extraction techniques. The process, called the ARALEX process, uses 2-ethyl-1-hexanol (2-EHOH) to extract the TBP degradation products (HDBP and H 2 MBP) from acidified Na 2 CO 3 scrub waste leaving the actinides in the aqueous phase. Dibutyl and monobutyl phosphoric acids are attached to the 2-EHOH molecules through hydrogen bonds, which also diminish the ability of the HDBP and H 2 MBP to complex actinides. Thus all actinides remain in the aqueous raffinate. Dilute sodium hydroxide solutions can be used to back-extract the dibutyl and monobutyl phosphoric acid esters as their sodium salts. The 2-EHOH can then be recycled. After extraction of the acidified carbonate waste with 2-EHOH, the actinides may be readily extracted from the raffinate with DHDECMP or, in the case of tetra- and hexavalent actinides, with TBP. The ARALEX process can also be applied to other actinide waste streams which contain appreciable concentrations of polar organic compounds (e.g., detergents) that interfere with conventional actinide ion exchange and liquid-liquid extraction procedures. 20 figures, 6 tables

Advanced waste forms from spent nuclear fuel

International Nuclear Information System (INIS)

Ackerman, J.P.; McPheeters, C.C.

1995-01-01

More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed

Removal of actinides from high-level wastes generated in the reprocessing of commercial fuels

International Nuclear Information System (INIS)

Bond, W.D.; Leuze, R.E.

1975-09-01

Progress is reported on a technical feasibility study of removing the very long-lived actinides (uranium, neptunium, plutonium, americium, and curium) from high-level wastes generated in the commercial reprocessing of spent nuclear fuels. The study was directed primarily at wastes from the reprocessing of light water reactor (LWR) fuels and specifically to developing satisfactory methods for reducing the actinide content of these wastes to values that would make 1000-year-decayed waste comparable in radiological toxicity to natural uranium ore deposits. Although studies are not complete, results thus far indicate the most promising concept for actinide removal includes both improved recovery of actinides in conventional fuel reprocessing and secondary processing of the high-level wastes. Secondary processing will be necessary for the removal of americium and curium and perhaps some residual plutonium. Laboratory-scale studies of separations methods that appear most promising are reported and conceptual flowsheets are discussed. (U.S.)

Polyphase ceramic and glass-ceramic forms for immobilizing ICPP high-level nuclear waste

International Nuclear Information System (INIS)

Harker, A.B.; Flintoff, J.F.

1984-01-01

Polyphase ceramic and glass-ceramic forms have been consolidated from simulated Idaho Chemical Processing Plant wastes by hot isostatic pressing calcined waste and chemical additives by 1000 0 C or less. The ceramic forms can contain over 70 wt% waste with densities ranging from 3.5 to 3.85 g/cm 3 , depending upon the formulation. Major phases are CaF 2 , CaZrTi 207 , CaTiO 3 , monoclinic ZrO 2 , and amorphous intergranular material. The relative fraction of the phases is a function of the chemical additives (TiO 2 , CaO, and SiO 2 ) and consolidation temperature. Zirconolite, the major actinide host, makes the ceramic forms extremely leach resistant for the actinide simulant U 238 . The amorphous phase controls the leach performance for Sr and Cs which is improved by the addition of SiO 2 . Glass-ceramic forms were also consolidated by HIP at waste loadings of 30 to 70 wt% with densities of 2.73 to 3.1 g/cm 3 using Exxon 127 borosilicate glass frit. The glass-ceramic forms contain crystalline CaF 2 , Al 203 , and ZrSi 04 (zircon) in a glass matrix. Natural mineral zircon is a stable host for 4+ valent actinides. 17 references, 3 figures, 5 tables

Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

International Nuclear Information System (INIS)

Gompper, K.

1986-01-01

The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

Overall assessment of actinide partitioning and transmutation for waste management purposes

International Nuclear Information System (INIS)

Blomeke, J.O.; Croff, A.G.; Finney, B.C.; Tedder, D.W.

1980-01-01

A program to establish the technical feasibility and incentives for partitioning (i.e., recovering) actinides from fuel cycle wastes and then transmuting them in power reactors to shorter-lived or stable nuclides has recently been concluded at the Oak Ridge National Laboratory. The feasibility was established by experimentally investigating the reduction that can be practicably achieved in the actinide content of the wastes sent to a geologic repository, and the incentives for implementing this concept were defined by determining the incremental costs, risks, and benefits. Eight US Department of Energy laboratories and three private companies participated in the program over its 3-year duration. A reference fuel cycle was chosen based on a self-generated plutonium recycle PWR, and chemical flowsheets based on solvent extraction and ion-exchange techniques were generated that have the potential to reduce actinides in fuel fabrication and reprocessing plant wastes to less than 0.25% of those in the spent fuel. Waste treatment facilities utilizing these flowsheets were designed conceptually, and their costs were estimated. Finally, the short-term (contemporary) risks from fuel cycle operations and long-term (future) risks from deep geologic disposal of the wastes were estimated for cases with and without partitioning and transmutation. It was concluded that, while both actinide partitioning from wastes and transmutation in power reactors appear to be feasible using currently identified and studied technology, implementation of this concept cannot be justified because of the small long-term benefits and substantially increased costs of the concept

Method for the recovery of actinide elements from nuclear reactor waste

International Nuclear Information System (INIS)

Horwitz, E.P.; Delphin, W.H.; Mason, G.W.

1979-01-01

A process is described for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of dihexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values. Uranium values remain in the extractant and may be recovered with a phosphoric acid strip. The neptunium and plutonium values are recovered from the oxalate by adding sufficient nitric acid to destroy the complexing ability of the oxalate, forming a second feed, and contacting the second feed with a second extractant of tricaprylmethylammonium nitrate in an inert diluent whereby the neptunium and plutonium values are selectively extracted. The values are recovered from the extractant with formic acid

Actinide partitioning from high level liquid waste using the Diamex process

International Nuclear Information System (INIS)

Madic, C.; Blanc, P.; Condamines, N.; Baron, P.; Berthon, L.; Nicol, C.; Pozo, C.; Lecomte, M.; Philippe, M.; Masson, M.; Hequet, C.

1994-01-01

The removal of long-lived radionuclides, which belong to the so-called minor actinides elements, neptunium, americium and curium, from the high level nuclear wastes separated during the reprocessing of the irradiated nuclear fuels in order to transmute them into short-lived nuclides, can substantially decrease the potential hazards associated with the management of these nuclear wastes. In order to separate minor actinides from high-level liquid wastes (HLLW), a liquid-liquid extraction process was considered, based on the use of diamide molecules, which display the property of being totally burnable, thus they do not generate secondary solid wastes. The main extracting properties of dimethyldibutyltetradecylmalonamide (DMDBTDMA), the diamide selected for the development of the DIAMEX process, are briefly described in this paper. Hot tests of the DIAMEX process (using DMDBTDMA) related to the treatment of an mixed oxide fuels (MOX) type HLLW, were successfully performed. The minor actinide decontamination factors of the HLLW obtained were encouraging. The main results of these tests are presented and discussed in this paper. (authors). 9 refs., 2 figs., 7 tabs

Elimination of waste actinides by recycling them to nuclear reactors

International Nuclear Information System (INIS)

McKay, H.A.C.

1981-01-01

After a few centuries of radioactive decay the long-lived actinides, the elements of atomic numbers 89-103, may constitute the main potential radiological health hazard in nuclear wastes. This is because all but a very few fission products (principally technetium-99 and iodine-129) have by then decayed to insignificant levels, leaving the actinides as the principal hazardous species remaining. It is therefore at first sight an attractive idea to recycle the actinides in nuclear reactors, so as to eliminate them by nuclear fission. There are good reasons for examining the idea in detail, and studies have been carried out in a number of countries. These have culminated recently in international conferences at the European Joint Research Centre at Ispra in Italy and at Austin, Texas in the USA as well as in the issue of an IAEA Technical Report entitled An Evaluation of Actinide Partitioning and Transmutation, a product of a four-year IAEA Co-ordinated Research Programme, on which the present article is based. The term partitioning refers to the separation of the actinides from nuclear fuel cycle wastes, a necessary preliminary step to their introduction into reactors for transmutation by nuclear fission. The complete scheme will be referred to as P-T, i.e. partitioning-transmutation

Transmutation of fission products and actinide waste at Hanford

Energy Technology Data Exchange (ETDEWEB)

Daemen, L.L.; Pitcher, E.J.; Russell, G.J. [Los Alamos National Laboratory, NM (United States)

1995-10-01

The authors studied the neutronics of an ATW system for the transmutation of the fission products ({sup 99}Tc in particular) and the type of actinide waste stored in several tanks at Hanford. The heart of the system is a highly-efficient neutron production target. It is surrounded by a blanket containing a moderator/reflector material, as well as the products to be transmuted. The fission products are injected into the blanket in the form of an aqueous solution in heavy water, whereas an aqueous actinides slurry is circulated in the outer part of the blanket. For the sake of definiteness, the authors focussed on {sup 99}Tc (the most difficult fission product to transmute), and {sup 239}Pu, {sup 237}Np, and {sup 241}Am. Because of the low thermal neutron absorption cross-section of {sup 99}Tc, considerable care and effort must be devoted to the design of a very efficient neutron source.

Characterization Of Actinides In Simulated Alkaline Tank Waste Sludges And Leachates

International Nuclear Information System (INIS)

Nash, Kenneth L.

2008-01-01

In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

CHARACTERIZATION OF ACTINIDES IN SIMULATED ALKALINE TANK WASTE SLUDGES AND LEACHATES

Energy Technology Data Exchange (ETDEWEB)

Nash, Kenneth L.

2008-11-20

In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

Hydrothermal processing of actinide contaminated organic wastes

International Nuclear Information System (INIS)

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

1997-01-01

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

Actinide-Aluminate Speciation in Alkaline Radioactive Waste

International Nuclear Information System (INIS)

Clark, David L.; Fedosseev, Alexander M.

2001-01-01

Investigation of behavior of actinides in alkaline media containing AL(III) showed that no aluminate complexes of actinides in oxidation states (IIII-VIII) were formed in alkaline solutions. At alkaline precipitation IPH (10-14) of actinides in presence of AL(III) formation of aluminate compounds is not observed. However, in precipitates contained actinides (IIV)<(VI), and to a lesser degree actinides (III), some interference of components takes place that is reflected in change of solid phase properties in comparison with pure components or their mechanical mixture. The interference decreases with rise of precipitation PH and at PH 14 is exhibited very feebly. In the case of NP(VII) the individual compound with AL(III) is obtained, however it is not aluminate of neptunium(VII), but neptunate of aluminium(III) similar to neptunates of other metals obtained earlier

Metal waste forms from treatment of EBR-II spent fuel

International Nuclear Information System (INIS)

Abraham, D. P.

1998-01-01

Demonstration of Argonne National Laboratory's electrometallurgical treatment of spent nuclear fuel is currently being conducted on irradiated, metallic driver fuel and blanket fuel elements from the Experimental Breeder Reactor-II (EBR-II) in Idaho. The residual metallic material from the electrometallurgical treatment process is consolidated into an ingot, the metal waste form (MWF), by employing an induction furnace in a hot cell. Scanning electron microscopy (SEM) and chemical analyses have been performed on irradiated cladding hulls from the driver fuel, and on samples from the alloy ingots. This paper presents the microstructures of the radioactive ingots and compares them with observations on simulated waste forms prepared using non-irradiated material. These simulated waste forms have the baseline composition of stainless steel - 15 wt % zirconium (SS-15Zr). Additions of noble metal elements, which serve as surrogates for fission products, and actinides are made to that baseline composition. The partitioning of noble metal and actinide elements into alloy phases and the role of zirconium for incorporating these elements is discussed in this paper

Challenges in Modeling the Degradation of Ceramic Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Devanathan, Ramaswami; Gao, Fei; Sun, Xin

2011-09-01

We identify the state of the art, gaps in current understanding, and key research needs in the area of modeling the long-term degradation of ceramic waste forms for nuclear waste disposition. The directed purpose of this report is to define a roadmap for Waste IPSC needs to extend capabilities of waste degradation to ceramic waste forms, which overlaps with the needs of the subconsinuum scale of FMM interests. The key knowledge gaps are in the areas of (i) methodology for developing reliable interatomic potentials to model the complex atomic-level interactions in waste forms; (ii) characterization of water interactions at ceramic surfaces and interfaces; and (iii) extension of atomic-level insights to the long time and distance scales relevant to the problem of actinide and fission product immobilization.

Challenges in Modeling the Degradation of Ceramic Waste Forms

International Nuclear Information System (INIS)

Devanathan, Ramaswami; Gao, Fei; Sun, Xin

2011-01-01

We identify the state of the art, gaps in current understanding, and key research needs in the area of modeling the long-term degradation of ceramic waste forms for nuclear waste disposition. The directed purpose of this report is to define a roadmap for Waste IPSC needs to extend capabilities of waste degradation to ceramic waste forms, which overlaps with the needs of the subconsinuum scale of FMM interests. The key knowledge gaps are in the areas of (i) methodology for developing reliable interatomic potentials to model the complex atomic-level interactions in waste forms; (ii) characterization of water interactions at ceramic surfaces and interfaces; and (iii) extension of atomic-level insights to the long time and distance scales relevant to the problem of actinide and fission product immobilization.

Method for the recovery of actinide elements from nuclear reactor waste

Science.gov (United States)

Horwitz, E. Philip; Delphin, Walter H.; Mason, George W.

1979-01-01

A process for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of dihexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate solution of tetramethylammonium hydrogen oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values. Uranium values remain in the extractant and may be recovered with a phosphoric acid strip. The neptunium and plutonium values are recovered from the oxalate by adding sufficient nitric acid to destroy the complexing ability of the oxalate, forming a second feed, and contacting the second feed with a second extractant of tricaprylmethylammonium nitrate in an inert diluent whereby the neptunium and plutonium values are selectively extracted. The values are recovered from the extractant with formic acid.

Removal of actinides from high activity wastes by solvent extraction: outline of the research work at Ispra J.R.C. laboratories

International Nuclear Information System (INIS)

Mannone, F.

1976-07-01

The development of an advanced waste management alternative such as the actinide nuclear incineration requires an almost quantitative removal of actinides from waste streams. Within the framework of the Ispra JRC Waste Disposal R and D programme, actinide separation studies were directed towards solvent extraction and precipitation methods. To develop a tentative waste partitioning flow-sheet based on solvent extraction, two conceptual process flow-sheet for actinide removal were evaluated on the basis of the currently used actinide recovery processes, i.e. removal after waste adjustment to low-acidity conditions and direct actinide removal from acidic wastes, as they are generated in actual reprocessing plants. No improvements have been devised for actinide recoveries within the conventional Purex reprocessing operations and a currently agreed value has been assumed for neptunium recovery (90%). According to these basic orientations some organic extractants have been selected for testing as promising candidates for waste partitioning and laboratory studies, designed to develop a satisfactory partitioning flow-sheet, have been proposed and described

ANSTO's waste forms for the 31. century

Energy Technology Data Exchange (ETDEWEB)

Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

2004-07-01

ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

International Nuclear Information System (INIS)

Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D.; Scales, Charlie R.; Maddrell, Ewan R.; Hobbs, Jeff

2013-01-01

A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

Removal of actinides from nuclear fuel reprocessing waste solutions with bidentate organophosphorus extractants

International Nuclear Information System (INIS)

Schulz, W.W.; McIsaac, L.D.

1975-08-01

The neutral bidentate organophosphorus reagents DBDECMP (dibutyl-N,N-diethylcarbamylmethylenephosphonate) and its dihexyl analogue DHDECMP are candidate extractants for removal of actinides from certain acidic waste streams produced at the U. S. ERDA Hanford and Idaho Falls sites. Various chemical and physical properties including availability, cost, purification, alpha radiolysis, and aqueous phase solubility of DBDECMP and DHDECMP are reviewed. A conceptual flowsheet employing a 15 percent DBDECMP (or DHDECMP)--CCl 4 extractant for removal (and recovery) of Am and Pu from Hanford's Plutonium Reclamation Facility acid waste stream (CAW solution) was successfully demonstrated in laboratory-scale mixer-settler tests; this extraction scheme can be used to produce an actinide-free waste. A 30 percent DBDECMP-xylene flowsheet is being tested at the Idaho Falls site for removal of U, Np, Pu, and Am from Idaho Chemical Processing Plant first-cycle high-level raffinate to produce an actinide-free (less than 10 nCi alpha activity/gram) waste. (auth)

Interaction of selected actinides (U, Cm) with bacteria relevant to nuclear waste disposal

Energy Technology Data Exchange (ETDEWEB)

Luetke, Laura

2013-07-01

To assess the safety of a site destined for storage of nuclear waste enhanced research effort is demanded to investigate the complex interactions of released radionuclides with parts of the environment that includes indigenous microorganisms. This work aimed at assessing the interactions of two bacterial strains with the actinides uranium and curium with a focus on thermodynamics to provide stability constants of the actinide bacteria species formed usable for modelling the distribution of these actinides in the environment. The influences of Pseudomonas fluorescens (CCUG 32456A) isolated from the granitic aquifers at Aespoe(Sweden) and a novel isolate from Mont Terri Opalinus clay (Switzerland), Paenibacillus sp. MT-2.2, were investigated. A combined approach using microbiological and spectroscopic techniques as well as potentiometry was employed to characterize the U(VI) and Cm(III) binding onto the cell surface functional groups structurally and thermodynamically. Further, due to its similar ionic radius to Cm(III) also Eu(III) was studied as non-radioactive analog.

Legal and regulatory issues regarding classification and disposal of wastes from actinide partitioning and transmutation

International Nuclear Information System (INIS)

Kocher, D.C.

1989-01-01

Partitioning and transmutation of actinide radioelements in spent nuclear fuel from civilian power reactors is potentially attractive because the resulting wastes might be acceptable for disposal using systems which are considerably less costly than a deep geologic repository. At present, there are no legal or regulatory prohibitions to seeking alternatives to a geologic repository for disposal of such wastes. However, additional laws and regulations would be needed, and the Nuclear Regulatory Commission has been reluctant to alter the current framework for radioactive waste management, in which geologic repositories or near-surface facilities are the only disposal options established in law and regulations unless a compelling need for alternatives with intermediate waste-isolation capabilities is demonstrated. There are also important technical considerations which are not encouraging with regard to the development of intermediate disposal systems for wastes from partitioning and transmutation of actinides in civilian spent fuel. First, the wastes may contain sufficient concentrations of fission products. Second, defense reprocessing wastes may contain sufficient concentrations of fission products and long-lived actinides. Thus, in developing the legal and regulatory framework for alternative disposal systems, there is a need to establish maximum concentrations of fission products and long-lived actinides that would be acceptable for intermediate disposal. 19 refs

Characterization of high level waste for minor actinides by chemical separation and alpha spectrometry

International Nuclear Information System (INIS)

Murali, M.S.; Bhattacharayya, A.; Kar, A.S.; Tomar, B.S.; Manchanda, V.K.

2010-01-01

Quantification of minor actinides present in of High Level Waste (HLW) solutions originating from the power reactors is important in view of management of radioactive wastes and actinide partitioning. Several methods such as ICP-MS, X-ray fluorescence methods, ICP-AES, alpha spectrometry are used in characterizing such types of wastes. As alpha spectrometry is simple and reliable, this technique has been used for the estimation of minor actinides after devising steps of separation for estimating Np and Pu present in HLW solutions of PHWR origin. Using a wealth of knowledge appropriate to the solution chemistry of actinides, the task of separation, though appears easy, it is challenging job for a radiochemist handling high-dose HLW samples, for obtaining clean alpha peaks for Np and Pu. This paper reports on the successful attempt made to quantify 241 Am, 244 Cm, Pu (239 mainly) and 237 Np present in HLW-PHWR obtained from PREFRE, Tarapur

Review of the sorption of actinides on natural minerals

International Nuclear Information System (INIS)

Beall, G.W.

1981-01-01

Over the past few years, a large body of data concerning sorption of actinides on geologic media has been built in connection with high-level-waste disposal. The primary aim of the work has been to allow predictions of the migration behavior of these radionuclides in the case of a breach of the repository that allowed groundwater flow through the repository. As a result of this work, some new backfill materials specifically tailored for the actinides have also been designed. Several major mechanisms of sorption that appear to dominate the sorption of actinides have emerged from these studies. These mechanisms can be divided into solution reactions dominated by hydrolysis, chemisorption reactions, and oxidation-reduction reactions. Each of these mechanisms will be discussed in detail, with experimental examples. Surprisingly, one mechanism, cation exchange, does not play an important role; why it fails to operate in any significant way in the environmental pH region will be discussed. The implications of the sorption mechanisms for waste forms and backfill materials will be discussed in detail. These discussions will center primarily around the valence state of the actinide in various waste forms and the effect of various anions on leachability from waste forms and backfill materials

Actinide separation by electrorefining

International Nuclear Information System (INIS)

Fusselman, S.P.; Gay, R.L.; Grantham, L.F.; Grimmett, D.L.; Roy, J.J.; Inoue, T.; Hijikata, T.; Krueger, C.L.; Storvick, T.S.; Takahashi, N.

1995-01-01

TRUMP-S is a pyrochemical process being developed for the recovery of actinides from PUREX wastes. This paper describes development of the electrochemical partitioning step for recovery of actinides in the TRUMP-S process. The objectives are to remove 99 % of each actinide from PUREX wastes, with a product that is > 90 % actinides. Laboratory tests indicate that > 99 % of actinides can be removed in the electrochemical partitioning step. A dynamic (not equilibrium) process model predicts that 90 wt % product actinide content can be achieved through 99 % actinide removal. Accuracy of model simulation results were confirmed in tests with rare earths. (authors)

Effects of actinide burning on waste disposal at Yucca Mountain

International Nuclear Information System (INIS)

Hirschfelder, J.

1992-01-01

Release rates of 15 radionuclides from waste packages expected to result from partitioning and transmutation of Light-Water Reactor (LWR) and Actinide-Burning Liquid-Metal Reactor (ALMR) spent fuel are calculated and compared to release rates from standard LWR spent fuel packages. The release rates are input to a model for radionuclide transport from the proposed geologic repository at Yucca Mountain to the water table. Discharge rates at the water table are calculated and used in a model for transport to the accessible environment, defined to be five kilometers from the repository edge. Concentrations and dose rates at the accessible environment from spent fuel and wastes from reprocessing, with partitioning and transmutation, are calculated. Partitioning and transmutation of LWR and ALMR spent fuel reduces the inventories of uranium, neptunium, plutonium, americium and curium in the high-level waste by factors of 40 to 500. However, because release rates of all of the actinides except curium are limited by solubility and are independent of package inventory, they are not reduced correspondingly. Only for curium is the repository release rate much lower for reprocessing wastes

Process and equipment qualification of the ceramic and metal waste forms for spent fuel treatment

International Nuclear Information System (INIS)

Marsden, Ken; Knight, Collin; Bateman, Kenneth; Westphal, Brian; Lind, Paul

2005-01-01

The electrometallurgical process for treating sodium-bonded spent metallic fuel at the Materials and Fuels Complex of the Idaho National Laboratory separates actinides and partitions fission products into two waste forms. The first is the metal waste form, which is primarily composed of stainless steel from the fuel cladding. This stainless steel is alloyed with 15w% zirconium to produce a very corrosion-resistant metal which binds noble metal fission products and residual actinides. The second is the ceramic waste form which stabilizes fission product-loaded chloride salts in a sodalite and glass composite. These two waste forms will be packaged together for disposal at the Yucca Mountain repository. Two production-scale metal waste furnaces have been constructed. The first is in a large argon-atmosphere glovebox and has been used for equipment qualification, process development, and process qualification - the demonstration of process reliability for production of the DOE-qualified metal waste form. The second furnace will be transferred into a hot cell for production of metal waste. Prototype production-scale ceramic waste equipment has been constructed or procured; some equipment has been qualified with fission product-loaded salt in the hot cell. Qualification of the remaining equipment with surrogate materials is underway. (author)

Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

Energy Technology Data Exchange (ETDEWEB)

Steven Frank; Hwan Seo Park; Yung Zun Cho; William Ebert; Brian Riley

2015-07-01

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.

Transmutation of waste actinides in thermal reactors: survey calculations of candidate irradiation schemes

International Nuclear Information System (INIS)

Gorrell, T.C.

1978-11-01

Actinide recycle and transmutation calculations were made for twelve specific thermal reactor environments. The calculations included H 2 O-moderated reactor lattices with enriched U, recycled Pu, and 233 ' 235 U-Th. In addition two D 2 O reactor cases were calculated. When all actinides were recycled into 235 U-enriched fuel, about 10 percent of the transuranic actinides were fissioned per 3-year fuel cycle. About 9 percent of the actinides were fissioned per 3-year fuel cycle when waste actinides (no U or Pu) were irradiated in separate target rods in a U-fuel assembly. When actinides were recycled in separate target assemblies, the fission rate was strongly dependent on the specific loading of the target. Fission rates of 5 to 10 percent per 3-year fuel cycle were observed

Actinide removal from wastewater applying waste minimization techniques

International Nuclear Information System (INIS)

Covey, J.R.; Midkiff, W.S.; Cadena, F.

1992-01-01

A major concern at LANL is the large volume of low level radioactive sludge that is generated by the current treatment technology. The plant meets current discharge limits but annually produces 200 55-gallon drums of sludge (approximately 60 tons) during the process of removing only few grants of radioactive isotopes. Most of the sludge results from the coagulants, iron and lime, added at the plant at a concentration of 10,000 parts-per-million (ppm). If the principal actinides in the influent could be separated and reduced to pure metallic form, the annual volume of plutonium would be about the size of a marble and the americium would be about the size of a BB. Waste minimization will be a key design criteria for the new facility. Records of total suspended solids (TSS) in the influent average about 1000 Kg per year (approximately 1 ton). Therefore, the theoretical sludge volume reduction is near 98%. Research is underway to develop and evaluate technologies that achieve the desired removal efficiency with a minimum of produced waste volume

Actinide speciation in the environment

International Nuclear Information System (INIS)

Choppin, G.R.

2007-01-01

Nuclear test explosions and nuclear reactor wastes and accidents have released large amounts of radioactivity into the environment. Actinide ions in waters often are not in a state of thermodynamic equilibrium and their solubility and migration behavior is related to the form in which the nuclides are introduced into the aquatic system. Chemical speciation, oxidation state, redox reactions, and sorption characteristics are necessary in predicting solubility of the different actinides, their migration behaviors and their potential effects on marine biota. The most significant of these variables is the oxidation state of the metal ion as the simultaneous presence of more than one oxidation state for some actinides in a solution complicates actinide environmental behavior. Both Np(V)O 2 + and Pu(V)O 2 + , the most significant soluble states in natural oxic waters, are relatively noncomplexing and resistant to hydrolysis and subsequent precipitation. The solubility of NpO 2 + can be as high as 10 -4 M while that of PuO 2 + is much more limited by reduction to the insoluble tetravalent species, Pu(OH) 4 , (pK sp ≥56) but which can be present in the pentavalent form in aqautic phases as colloidal material. The solubility of hexavalent UO 2 2+ in sea water is relatively high due to formation of carbonate complexes. The insoluble trivalent americium hydroxocarbonate, Am(OH)(CO 3 ) is the limiting species for the solubility of Am(III) in sea water. Thorium(IV) is present as Th(OH) 4 , in colloidal form. The chemistry of actinide ions in the environment is reviewed to show the spectrum of reactions that can occur in natural waters which must be considered in assessing the environmental behavior of actinides. Much is understood about sorption of actinides on surfaces, the mode of migration of actinides in such waters and the potential effects of these radioactive species on marine biota, but much more understanding of the behavior of the actinides in the environment is

Some activities in the United States concerning the physics aspects of actinide waste recycling

International Nuclear Information System (INIS)

Raman, S.

1975-01-01

Reactor types being considered in the United States for the purpose of actinide waste recycling are discussed briefly. The reactor types include thermal reactors operating on the 3.3 percent 235 U-- 238 U and the 233 U-- 232 Th fuel cycles, liquid metal fast breeder reactors, reactors fueled entirely by actinide wastes, gaseous fuel reactors, and fusion reactors. Cross section measurements in progress or planned toward providing basic data for testing the recycle concept are also discussed

Microbial Transformations of Actinides and Fission Products in Radioactive Waste

Energy Technology Data Exchange (ETDEWEB)

Francis, A. J. [Pohang Univ. Science and Technology, Pohang (Korea, Republic of)

2011-07-01

The environmental factors that can affect microbial growth and activity include moisture, temperature, ph, Eh, availability of organic and inorganic nutrients, and radiation. The microbial activity in a specific repository is influenced by the ambient environment of the repository, and the materials to be emplaced. For example, a repository in unsaturated igneous rock formations such as volcanic tuff rocks at Yucca Mountain is generally expected to be oxidizing; a repository in a hydrologically expected to be oxidizing; a repository in a hydrologically saturated zone, especially in sedimentary rocks, could be reducing. Sedimentary rocks contain a certain amount of organic matter, which may stimulate microbial activities and, thus maintain the repository and its surrounding areas at reducing conditions. Although the impacts of microbial activity on high-level nuclear waste and the long-term performance of the repository have not fully investigated, little microbial activity is expected in the near-field because of the radiation, lack of nutrients and the harsh conditions. However in the far-field microbial effects could be significant. Much of our understanding of the microbial effects on radionuclides stems from studies conducted with selected transuranic elements and fission products and limited studies with low-level radioactive wastes. Significant aerobic- and anaerobic-microbial activity is expected to occur in the waste because of the presence of electron donors and acceptors. The actinides initially may be present as soluble- or insoluble-forms but, after disposal, may be converted from one to the other by microorganisms. The direct enzymatic or indirect non-enzymatic actions of microbes could alter the speciation, solubility, and sorption properties of the actinides, thereby increasing or decreasing their concentrations in solution.

Selective extraction of actinides from high level liquid wastes. Study of the possibilities offered by the Redox properties of actinides

International Nuclear Information System (INIS)

Adnet, J.M.

1991-07-01

Partitioning of high level liquid wastes coming from nuclear fuel reprocessing by the PUREX process, consists in the elimination of minor actinides (Np, Am, and traces of Pu and U). Among the possible processes, the selective extraction of actinides with oxidation states higher than three is studied. First part of this work deals with a preliminary step; the elimination of the ruthenium from fission products solutions using the electrovolatilization of the RuO4 compound. The second part of this work concerns the complexation and oxidation reactions of the elements U, Np, Pu and Am in presence of a compound belonging to the insaturated polyanions family: the potassium phosphotungstate. For actinide ions with oxidation state (IV) complexed with phosphotungstate anion the extraction mechanism by dioctylamine was studied and the use of a chromatographic extraction technic permitted successful separations between tetravalents actinides and trivalents actinides. Finally, in accordance with the obtained results, the basis of a separation scheme for the management of fission products solutions is proposed

Actinides and fission products partitioning from high level liquid waste

International Nuclear Information System (INIS)

Yamaura, Mitiko

1999-01-01

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

Actinide cross section data and inertial confinement fusion for long term waste disposal

International Nuclear Information System (INIS)

Meldner, H.

1979-01-01

Actinide cross section data at thermonuclear neutron energies are needed for the calculation of ICF pellet center burnup of fission reactor waste, viz. 14 MeV neutron fission of the very long-lived actinides that pose storage problems. A major advantage of pellet center burnup is safety: only milligrams of highly toxic and active material need to be present in the fusion chamber, whereas blanket burnup requires the continued presence of tons of actinides in a small volume. The actinide data tables required for Monte Carlo calculations of the burnup of 241 Am and 243 Am are discussed in connection with typical burnup reactor fusion and fission spectra. 2 figures

Trends in actinide processing at Hanford

International Nuclear Information System (INIS)

Harmon, H.D.

1993-09-01

In 1989, the mission at the Hanford Site began a dramatic and sometimes painful transition. The days of production--as we used to know it--are over. Our mission officially has become waste management and environmental cleanup. This mission change didn't eliminate many jobs--in fact, budgets have grown dramatically to support the new mission. Most all of the same skilled crafts, engineers, and scientists are still required for the new mission. This change has not eliminated the need for actinide processing, but it has certainly changed the focus that our actinide chemists and process engineers have. The focus used to be on such things as increasing capacity, improving separations efficiency, and product purity. Minimizing waste had become a more important theme in recent years and it is still a very important concept in the waste management and environmental cleanup arena. However, at Hanford, a new set of words dominates the actinide process scene as we work to deal with actinides that still reside in a variety of forms at the Hanford Site. These words are repackage, stabilize, remove, store and dispose. Some key activities in each of these areas are described in this report

Some activities in the United States concerning the physics aspects of actinide waste recycling

International Nuclear Information System (INIS)

Raman, S.

1976-01-01

This review paper briefly discusses the reactor types being considered in the United States for the purpose of actinide waste recycling. The reactor types include thermal reactors operating on the 3.3% 235 U- 238 U and the 233 U- 232 Th fuel cycles, liquid metal fast breeder reactors, reactors fueled entirely by actinide wastes, gaseous fuel reactors and fusion reactors. This paper also discusses cross section measurements in progress or planned toward providing basic data for testing the recycle concept. (author)

Removal of actinides from dilute waste waters using polymer filtration

International Nuclear Information System (INIS)

Smith, B.F.; Robison, T.W.; Gibson, R.R.

1995-01-01

More stringent US Department of Energy discharge regulations for waste waters containing radionuclides (30 pCi/L total alpha) require the development of new processes to meet the new discharge limits for actinide metal ions, particularly americium and plutonium, while minimizing waste. We have been investigating a new technology, polymer filtration, that has the potential for effectively meeting these new limits. Traditional technology uses basic iron precipitation which produces large amounts of waste sludge. The new technology is based on using water-soluble chelating polymers with ultrafiltration for physical separation. The actinide metal ions are selectively bound to the polymer and can not pass through the membrane. Small molecules and nonbinding metals pass through the membrane. Advantages of polymer filtration technology compared to ion, exchange include rapid kinetics because the binding is occurring in a homogenous solution and no mechanical strength requirement on the polymer. We will present our results on the systematic development of a new class of water-soluble chelating polymers and their binding ability from dilute acid to near neutral waters

Minor actinide transmutation using minor actinide burner reactors

International Nuclear Information System (INIS)

Mukaiyama, T.; Yoshida, H.; Gunji, Y.

1991-01-01

The concept of minor actinide burner reactor is proposed as an efficient way to transmute long-lived minor actinides in order to ease the burden of high-level radioactive waste disposal problem. Conceptual design study of minor actinide burner reactors was performed to obtain a reactor model with very hard neutron spectrum and very high neutron flux in which minor actinides can be fissioned efficiently. Two models of burner reactors were obtained, one with metal fuel core and the other with particle fuel core. Minor actinide transmutation by the actinide burner reactors is compared with that by power reactors from both the reactor physics and fuel cycle facilities view point. (author)

Disposition of actinides released from high-level waste glass

International Nuclear Information System (INIS)

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

1994-01-01

A series of static leach tests was conducted using glasses developed for vitrifying tank wastes at the Savannah River Site to monitor the disposition of actinide elements upon corrosion of the glasses. In these tests, glasses produced from SRL 131 and SRL 202 frits were corroded at 90 degrees C in a tuff groundwater. Tests were conducted using crushed glass at different glass surface area-to-solution volume (S/V) ratios to assess the effect of the S/V on the solution chemistry, the corrosion of the glass, and the disposition of actinide elements. Observations regarding the effects of the S/V on the solution chemistry and the corrosion of the glass matrix have been reported previously. This paper highlights the solution analyses performed to assess how the S/V used in a static leach test affects the disposition of actinide elements between fractions that are suspended or dissolved in the solution, and retained by the altered glass or other materials

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Actinide Separation Demonstration Facility, Tarapur

International Nuclear Information System (INIS)

Vishwaraj, I.

2017-01-01

Partitioning of minor actinide from high level waste could have a substantial impact in lowering the radio toxicity associated with high level waste as well as it will reduce the burden on geological repository. In Indian context, the partitioned minor actinide could be routed into the fast breeder reactor systems scheduled for commissioning in the near period. The technological breakthrough in solvent development has catalyzed the partitioning programme in India, leading to the setting up and hot commissioning of the Actinide Separation Demonstration Facility (ASDF) at BARC, Tarapur. The engineering scale Actinide Separation Demonstration Facility (ASDF) has been retrofitted in an available radiological hot cell situated adjacent to the Advanced Vitrification Facility (AVS). This location advantage ensures an uninterrupted supply of high-level waste and facilitates the vitrification of the high-level waste after separation of minor actinides

Actinide recycling for reactor waste mass and radiotoxicity reduction

International Nuclear Information System (INIS)

Renard, A.; Maldague, T.; Pilate, S.; Journet, J.; Rome, M.; Harislur, A.; Vergnes, J.

1994-01-01

The long-term radiotoxicity of nuclear waste from a Light Water Reactor fuel is analyzed; it can be reduced by multiple recycling of actinides in fast reactors. The capabilities of a first recycling in the light water reactor itself are evaluated with regard to implications on reactor physics and core management. Two main options are compared with their penalties and efficiency

Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

International Nuclear Information System (INIS)

Deshingkar, D.S.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Nair, M.K.T.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Rao, M.K.; Mathur, J.N.; Murali, M.S.; Iyer, R.H.; Badheka, L.P.; Banerji, A.

1994-01-01

High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs

Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

Energy Technology Data Exchange (ETDEWEB)

Deshingkar, D S; Chitnis, R R; Wattal, P K; Theyyunni, T K; Nair, M K.T. [Bhabha Atomic Research Centre, Bombay (India). Process Engineering and Systems Development Div.; Ramanujam, A; Dhami, P S; Gopalakrishnan, V; Rao, M K [Bhabha Atomic Research Centre, Bombay (India). Fuel Reprocessing Group; Mathur, J N; Murali, M S; Iyer, R H [Bhabha Atomic Research Centre, Bombay (India). Radiochemistry Div.; Badheka, L P; Banerji, A [Bhabha Atomic Research Centre, Bombay (India). Bio-organic Div.

1994-12-31

High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs.

Studies on the properties of hard-spectrum, actinide fissioning reactors. Final report

International Nuclear Information System (INIS)

Nelson, J.B.; Prichard, A.W.; Schofield, P.E.; Robinson, A.H.; Spinrad, B.I.

1980-01-01

It is technically feasible to construct an operable (e.g., safe and stable) reactor to burn waste actinides rapidly. The heart of the concept is a driver core of EBR-II type, with a central radial target zone in which fuel elements, made entirely of waste actinides are exposed. This target fuel undergoes fission, as a result of which actinides are rapidly destroyed. Although the same result could be achieved in more conventionally designed LWR or LMFBR systems, the fast spectrum reactor does a much more efficient job, by virtue of the fact that in both LWR and LMFBR reactors, actinide fission is preceded by several captures before a fissile nuclide is formed. In the fast spectrum reactor that is called ABR (actinide burning reactor), these neutron captures are short-circuited

Integral migration and source-term experiments on cement and bitumen waste forms

International Nuclear Information System (INIS)

Ewart, F.T.; Howse, R.M.; Sharpe, B.M.; Smith, A.J.; Thomason, H.P.; Williams, S.J.; Young, M.

1986-01-01

This is the final report of a programme of research which formed a part of the CEC joint research project into radionuclide migration in the geosphere (MIRAGE). This study addressed the aspects of integral migration and source term. The integral migration experiment simulated, in the laboratory, the intrusion of water into the repository, the leaching of radionuclides from two intermediate-level waste-forms and the subsequent migration through the geosphere. The simulation consisted of a source of natural ground water which flowed over a sample of waste-form, at a controlled redox potential, and then through backfill and geological material packed in columns. The two waste forms used here were cemented waste from the WAK plant at Karlsruhe in the Federal Republic of Germany and bitumenized intermediate concentrates from the Marcoule plant in France. The soluble fission products such as caesium were rapidly released from the cemented waste but the actinides, and technetium in the reduced state, were retained in the waste-form. The released of all nuclides from the bitumenized waste was very low

Heterogeneous all actinide recycling in LWR all actinide cycle closure concept

International Nuclear Information System (INIS)

Tondinelli, Luciano

1980-01-01

A project for the elimination of transuranium elements (Waste Actinides, WA) by neutron transmutation is developed in a commercial BWR with U-Pu (Fuel Actinides, FA) recycle. The project is based on the All Actinide Cycle Closure concept: 1) closure of the 'back end' of the fuel cycle, U-Pu coprocessing, 2) waste actinide disposal by neutron transmutation. The reactor core consists of Pu-island fuel assemblies containing WAs in target pins. Two parallel reprocessing lines for FAs and WAs are provided. Mass balance, hazard measure, spontaneous activity during 10 recycles are calculated. Conclusions are: the reduction in All Actinide inventory achieved by Heterogeneous All Actinide Recycling is on the order of 83% after 10 recycles. The U235 enrichment needed for a constant end of cycle reactivity decreases for increasing number of recycles after the 4th recycle. A diffusion-burnup calculation of the pin power peak factors in the fuel assembly shows that design limits can be satisfied. A strong effort should be devoted to the solution of the problems related to high values of spontaneous emission by the target pins

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

International Nuclear Information System (INIS)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

Energy Technology Data Exchange (ETDEWEB)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

Burning actinides in very hard spectrum reactors

International Nuclear Information System (INIS)

Robinson, A.H.; Shirley, G.W.; Prichard, A.W.; Trapp, T.J.

1978-01-01

The major unresolved problem in the nuclear industry is the ultimate disposition of the waste products of light water reactors. The study demonstrates the feasibility of designing a very hard spectrum actinide burner reactor (ABR). A 1100 MW/sub t/ ABR design fueled entirely with actinides reprocessed from light water reactor (LWR) wastes is proposed as both an ultimate disposal mechanism for actinides and a means of concurrently producing usable power. Actinides from discharged ABR fuel are recycled to the ABR while fission products are routed to a permanent repository. As an integral part of a large energy park, each such ABR would dispose of the waste actinides from 2 LWRs

Actinide phosphonate complexes in aqueous solutions

International Nuclear Information System (INIS)

Nash, K.L.

1993-01-01

Complexes formed by actinides with carboxylic acids, polycarboxylic acids, and aminopolycarboxylic acids play a central role in both the basic and process chemistry of the actinides. Recent studies of f-element complexes with phosphonic acid ligands indicate that new ligands incorporating doubly ionizable phosphonate groups (-PO 3 H 2 ) have many properties which are unique chemically, and promise more efficient separation processes for waste cleanup and environmental restoration. Simple diphosphonate ligands form much stronger complexes than isostructural carboxylates, often exhibiting higher solubility as well. In this manuscript recent studies of the thermodynamics and kinetics of f-element complexation by 1,1 and 1,2 diphosphonic acid ligands are described

The treatment and packaging of waste plutonium and waste actinides for disposal

International Nuclear Information System (INIS)

Taylor, R.F.

1988-07-01

The objectives of this work have been to review the current state of knowledge on the treatment and packaging of unusable or surplus plutonium and other waste actinides for disposal and to identify any gaps in data essential for the development of a preferred route. The exercise was based on published data which said the quantity currently to be disposed of was 50 tonnes in oxide form. A literature review over the period 1978 to 1988 was carried out and a computerised database specific to the exercise was created. From this it is concluded that there are no insuperable problems to the formulation of a disposal route although there is none currently proven. The preferred wasteform would be a glass or synthetic rock. The major complication lies in the fissile nature of plutonium which dictates limits to the package size and places restrictions on the production and disposal routes. Additional work necessary to permit a final decision is listed. (author)

Characterization of a glass-bonded ceramic waste form loaded with U and Pu

International Nuclear Information System (INIS)

Sinkler, W.; O'Holleran, T. P.; Frank, S. M.; Richmann, M. K.; Johnson, S. G.

1999-01-01

This paper presents microscopic characterization of four samples of a ceramic waste form (CWF) developed for disposal of actinide-containing electrorefiner salts. The four samples were prepared to investigate the influence of water content and the Pu:U ratio on CWF microstructure and performance. While the overall phase content is not strongly influenced by either variable, the presence of water in the initial zeolite has a detectable effect on CWF microstructure. It is found to influence the distribution of the major actinide host phase, a (U,Pu)O 2 mixed oxide

The reprocessing-recycling of spent nuclear fuel. Actinides separation - Application to wastes management

International Nuclear Information System (INIS)

2008-01-01

After its use in the reactor, the spent fuel still contains lot of recoverable material for an energetic use (uranium, plutonium), but also fission products and minor actinides which represent the residues of nuclear reactions. The reprocessing-recycling of the spent fuel, as it is performed in France, implies the chemical separation of these materials. The development and the industrial implementation of this separation process represent a major contribution of the French science and technology. The reprocessing-recycling allows a good management of nuclear wastes and a significant saving of fissile materials. With the recent spectacular rise of uranium prices, this process will become indispensable with the development of the next generation of fast neutron reactors. This book takes stock of the present and future variants of the chemical process used for the reprocessing of spent fuels. It describes the researches in progress and presents the stakes and recent results obtained by the CEA. content: the separation of actinides, a key factor for a sustainable nuclear energy; the actinides, a discovery of the 20. century; the radionuclides in nuclear fuels; the aquo ions of actinides; some redox properties of actinides; some complexing properties of actinide cations; general considerations about treatment processes; some characteristics of nuclear fuels in relation with their reprocessing; technical goals and specific constraints of the PUREX process; front-end operations of the PUREX process; separation and purification operations of the PUREX process; elaboration of finite products in the framework of the PUREX process; management and treatment of liquid effluents; solid wastes of the PUREX process; towards a joint management of uranium and plutonium: the COEX TM process; technical options of treatment and recycling techniques; the fuels of generation IV reactors; front-end treatment processes of advanced fuels; hydrometallurgical processes for future fuel cycles

Evaluation of actinide partitioning and transmutation

International Nuclear Information System (INIS)

1982-01-01

After a few centuries of radioactive decay the long-lived actinides, the elements of atomic numbers 89-103, may constitute the main potential radiological health hazard in nuclear wastes. This is because all but a very few fission products (principally technetium-99 and iodine-129) have by then undergone radioactive decay to insignificant levels, leaving the actinides as the principal radionuclides remaining. It was therefore at first sight an attractive concept to recycle the actinides to nuclear reactors, so as to eliminate them by nuclear fission. Thus, investigations of the feasibility and potential benefits and hazards of the concept of 'actinide partitioning and transmutation' were started in numerous countries in the mid-1970s. This final report summarizes the results and conclusions of technical studies performed in connection with a four-year IAEA Co-ordinated Research Programme, started in 1976, on the ''Environmental Evaluation and Hazard Assessment of the Separation of Actinides from Nuclear Wastes followed by either Transmutation or Separate Disposal''. Although many related studies are still continuing, e.g. on waste disposal, long-term safety assessments, and waste actinide management (particularly for low and intermediate-level wastes), some firm conclusions on the overall concept were drawn by the programme participants, which are reflected in this report

Actinide recycle

Energy Technology Data Exchange (ETDEWEB)

Till, C; Chang, Y [Argonne National Laboratory, Argonne, IL (United States)

1990-07-01

A multitude of studies and assessments of actinide partitioning and transmutation were carried out in the late 1970s and early 1980s. Probably the most comprehensive of these was a study coordinated by Oak Ridge National Laboratory. The conclusions of this study were that only rather weak economic and safety incentives existed for partitioning and transmuting the actinides for waste management purposes, due to the facts that (1) partitioning processes were complicated and expensive, and (2) the geologic repository was assumed to contain actinides for hundreds of thousands of years. Much has changed in the few years since then. A variety of developments now combine to warrant a renewed assessment of the actinide recycle. First of all, it has become increasingly difficult to provide to all parties the necessary assurance that the repository will contain essentially all radioactive materials until they have decayed. Assurance can almost certainly be provided to regulatory agencies by sound technical arguments, but it is difficult to convince the general public that the behavior of wastes stored in the ground can be modeled and predicted for even a few thousand years. From this point of view alone there would seem to be a clear benefit in reducing the long-term toxicity of the high-level wastes placed in the repository.

Actinide recycle

International Nuclear Information System (INIS)

Till, C.; Chang, Y.

1990-01-01

A multitude of studies and assessments of actinide partitioning and transmutation were carried out in the late 1970s and early 1980s. Probably the most comprehensive of these was a study coordinated by Oak Ridge National Laboratory. The conclusions of this study were that only rather weak economic and safety incentives existed for partitioning and transmuting the actinides for waste management purposes, due to the facts that (1) partitioning processes were complicated and expensive, and (2) the geologic repository was assumed to contain actinides for hundreds of thousands of years. Much has changed in the few years since then. A variety of developments now combine to warrant a renewed assessment of the actinide recycle. First of all, it has become increasingly difficult to provide to all parties the necessary assurance that the repository will contain essentially all radioactive materials until they have decayed. Assurance can almost certainly be provided to regulatory agencies by sound technical arguments, but it is difficult to convince the general public that the behavior of wastes stored in the ground can be modeled and predicted for even a few thousand years. From this point of view alone there would seem to be a clear benefit in reducing the long-term toxicity of the high-level wastes placed in the repository

Partitioning of actinides from high active waste solution of Purex origin counter-current extraction studies using TBP and CMPO

International Nuclear Information System (INIS)

Chitnis, R.R.; Dhami, P.S.; Gopalkrishnan, V.; Wattal, P.K.; Ramanujam, A.; Murali, M.S.; Mathur, J.N.; Bauri, A.K.; Chattopadhyay, S.

2000-10-01

A solvent extraction scheme has been formulated for the partitioning of actinides from Purex high level waste (HLW). The scheme is based on the results of earlier studies carried out with simulated waste solutions. In the present studies, the scheme was tested with high active waste (HAW) solution generated during the reprocessing of spent fuel from research reactors using laboratory scale mixer-settlers. The proposed process involved two-step extraction using tri-n-butyl phosphate (TBP) and octyl (phenyl)-N,N-diisobutylcarbamolylmethylphosphine oxide (CMPO). In the first step, uranium, neptunium and plutonium were removed from the waste using TBP as extractant. The minor actinides left in the raffinate were extracted using a mixture of CMPO and TBP in the second step. The results showed complete extraction of actinides from the waste solution. Plutonium and neptunium extracted in TBP, were stripped together using a mixture of hydrogen peroxide and ascorbic acid in 2 M nitric acid medium, leaving uranium in the organic phase. Uranium can later be stripped using dilute nitric acid. Actinides extracted in CMPO-TBP phase were stripped using a mixture of formic acid, hydrazine, hydrate and citric acid. The stripping was quantitative in both the stripping runs. An additional extraction step for the preferential recovery of uranium, neptunium and plutonium from the waste solution using TBP is a modification over the conventional Truex process. Selective stripping of neptunium and plutonium from large quantities of uranium. The extraction of uranium using TBP eliminates the possibility of third phase and undesired loading of CMPO-TBP in the following step. Use of citrate-containing strippant allows the recovery of actinides from loaded CMPO-TBP mixture without causing any reflux of the actinides during stripping. The process has been developed with due consideration to minimising the generation of secondary wastes. The proposed strippants are effective even in presence of

The effects of actinide separation on the radiological consequences of disposal of high-level radioactive waste on the ocean bed

International Nuclear Information System (INIS)

Camplin, W.C.; Grimwood, P.D.; White, I.F.

1980-01-01

One option in the management of high-level radioactive wastes is to separate the actinides prior to vitrification and disposal. This option is examined in the context of disposal of high-level wastes on the deep ocean bed. The initial quantity of waste corresponds to the generation of 1000 GW(e)y of nuclear energy, and the actinide-separation process is assumed to remove 99% of all elements of atomic number greater than that of actinium. The models used to describe the dispersion of activity from a single disposal site on the bed of the Atlantic Ocean represent both local dispersion and long-term mixing. Collective doses and doses to individuals are calculated for six potential pathways: ingestion of fish, crustacea, molluscs, plankton and seaweed, and external irradiation from contaminated beach sediments. The period from 400 to 1,000,000 years after disposal is considered. The potential radiological impact from disposal of high-level waste without separation of actinides on the ocean bed arises from the actinides; isotopes of americium, neptunium and plutonium give the highest doses. Actinide separation would reduce these doses in proportion to the effectiveness of the separation process, until doses become determined by fission products rather than actinides: the achievable dose reduction would be a factor of approximately a hundred, or less for certain pathways. This reduction applies only to doses to the public from waste disposal: no account was taken of doses arising from the separation process itself or from the management of the separated actinides. The results of the assessment are contrasted with those of similar studies based on toxicity indices. Major deficiencies are identified in the use of toxicity indices as a basis for decision-making. (author)

Separation of actinides and long-lived fission products from high-level radioactive wastes (a review)

International Nuclear Information System (INIS)

Kolarik, Z.

1991-11-01

The management of high-level radioactive wastes is facilitated, if long-lived and radiotoxic actinides and fission products are separated before the final disposal. Especially important is the separation of americium, curium, plutonium, neptunium, strontium, cesium and technetium. The separated nuclides can be deposited separately from the bulk of the high-level waste, but their transmutation to short-lived nuclides is a muchmore favourable option. This report reviews the chemistry of the separation of actinides and fission products from radioactive wastes. The composition, nature and conditioning of the wastes are described. The main attention is paid to the solvent extraction chemistry of the elements and to the application of solvent extraction in unit operations of potential partitioning processes. Also reviewed is the behaviour of the elements in the ion exchange chromatography, precipitation, electrolysis from aqueous solutions and melts, and the distribution between molten salts and metals. Flowsheets of selected partitioning processes are shown and general aspects of the waste partitioning are shortly discussed. (orig.) [de

New insights into phosphate based materials for the immobilisation of actinides

Energy Technology Data Exchange (ETDEWEB)

Neumeier, Stefan; Ji, Yaqi; Kowalski, Piotr M.; Kegler, Philip; Schlenz, Hartmut; Bosbach, Dirk; Deissmann, Guido [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK), Nukleare Entsorgung und Reaktorsicherheit (IEK-6); JARA High-Performance Computing, Aachen (Germany); Arinicheva, Yulia [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK), Nukleare Entsorgung und Reaktorsicherheit (IEK-6); JARA High-Performance Computing, Aachen (Germany); Forschungszentrum Juelich (Germany). Institut fuer Energie- und Klimaforschung (IEK), Werkstoffsynthese und Herstellungsverfahren (IEK-1); Heuser, Julia M. [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK), Nukleare Entsorgung und Reaktorsicherheit (IEK-6); JARA High-Performance Computing, Aachen (Germany); Karlsruhe Institute of Technology (Germany). Inst. of Applied Materials (IAM)

2017-07-01

This paper focuses on major phosphate-based ceramic materials relevant for the immobilisation of Pu, minor actinides, fission and activation products. Key points addressed include the recent progress regarding synthesis methods, the formation of solid solutions by structural incorporation of actinides or their non-radioactive surrogates and waste form fabrication by advanced sintering techniques. Particular attention is paid to the properties that govern the long-term stability of the waste forms under conditions relevant to geological disposal. The paper highlights the benefits gained from synergies of state-of-the-art experimental approaches and advanced atomistic modeling tools for addressing properties and stability of f-element-bearing phosphate materials. In conclusion, this article provides a perspective on the recent advancements in the understanding of phosphate based ceramics and their properties with respect to their application as nuclear waste forms.

Removal of actinide elements from high level radioactive waste by trialkylphosphine oxide (TRPO)

International Nuclear Information System (INIS)

Song Chongli; Yang Dazhu; He Longhai; Xu Jingming; Zhu Yongjun

1992-03-01

The modified TRPO process for removing actinide elements from synthetic solution, which was taken from reprocessing of power reactor nuclear fuel, was verified by cascade experiment. Neptunium valence was adjusted in the process for improving neptunium removing efficiency. At 1 mol/L concentration of HNO 3 of feed solution and after a few stages of extraction with 30% t=TRPO kerosene, over 99.9% of Am, Pu, Np and U could be removed from HAW (high level radioactive waste) solution. The stripping of actinides loaded in TRPO are accomplished by high concentration nitric acid, oxalic acid and sodium carbonate instead of amino carboxylic complexing agents used in previous process. The actinides stripped were divided into three groups, which are Am + RE, Np + Pu, and U, and the cross contamination between them is small. Behaviours of F.P. elements are divided into three types which are not extracted, little extracted and extracted elements. The extracted elements are rare earth and Pd, Zr and Mo which are co-extracted with actinides. The separation factor between actinides and other two types of F.P.elements will increase if more scrubbing sections are added in the process. The relative concentration profile of actinide elements and Tc in various stages as well as the distribution of actinides and F.P. elements in the process stream solutions are also presented

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

Demonstration of pyropartitioning process by using genuine high-level liquid waste. Reductive-extraction of actinide elements from chlorination product

International Nuclear Information System (INIS)

Uozumi, Koichi; Iizuka, Masatoshi; Kurata, Masaki; Ougier, Michel; Malmbeck, Rikard; Winckel, Stefaan van

2009-01-01

The pyropartitioning process separates the minor actinide elements (MAs) together with uranium and plutonium from the high-level liquid waste generated at the Purex reprocessing of spent LWR fuel and introduces them to metallic fuel cycle. For the demonstration of this technology, a series experiment using 520g of genuine high-level liquid waste was started and the conversion of actinide elements to their chlorides was already demonstrated by denitration and chlorination. In the present study, a reductive extraction experiment in molten salt/liquid cadmium system to recover actinide elements from the chlorination product of the genuine high-level liquid waste was performed. The results of the experiment are as following; 1) By the addition of the cadmium-lithium alloy reductant, almost all of plutonium and MAs in the initial high-level liquid waste were recovered in the cadmium phase. It means no mass loss during denitration, chlorination, and reductive-extraction. 2) The separation factor values of plutonium, MAs, and rare-earth fission product elements versus uranium agreed with the literature values. Therefore, actinide elements will be separated from fission product elements in the actual system. Hence, the pyropartitioning process was successfully demonstrated. (author)

Actinide recovery techniques utilizing electromechanical processes

International Nuclear Information System (INIS)

Westphal, B.R.; Benedict, R.W.

1994-01-01

Under certain conditions, the separation of actinides using electromechanical techniques may be an effective means of residue processing. The separation of granular mixtures of actinides and other materials is based on appreciable differences in the magnetic and electrical properties of the actinide elements. In addition, the high density of actinides, particularly uranium and plutonium, may render a simultaneous separation based on mutually complementary parameters. Both high intensity magnetic separation and electrostatic separation have been investigated for the concentration of an actinide waste stream. Waste stream constituents include an actinide metal alloy and broken quartz shards. The investigation of these techniques is in support of the Integral Fast Reactor (IFR) concept currently being developed at Argonne National Laboratory under the auspices of the Department of Energy

Actinide speciation bound to hydrous ferric oxide colloids in the near-field conditions of the waste pond at 'Mayak' facility (Russia)

International Nuclear Information System (INIS)

Kalmykov, St.; Khasanova, A.; Kriventsov, V.; Teterin, Y.; Novikov, A.

2007-01-01

Full text of publication follows: 'Mayak' facility is a nuclear waste and spent nuclear fuel reprocessing plant located in Ural Mountains, Russia. The opened pond, Karachay Lake, was used for several decades for the discharge of low- and intermediate level waste solutions containing fission products and traces of actinides. Due to high salt concentration and high density of waste solutions, they are penetrating into the groundwater system that is represented by oxic Eh conditions. The speciation of actinides in groundwater samples collected close to Karachay Lake was studied by successive micro- and ultra-filtrations with subsequent SEM, TEM, nano-SIMS, membrane extraction and other techniques. It was established that U and Np were found in soluble fraction (pass through 10 kD ultra-filter) in the form of their bi- and tri-carbonate complexes that was supported by chemical thermodynamic calculations. In contrast, Pu and Am were bound to nano-colloids 10 kD - 50 nm in size. The SEM and TEM data indicate the presence of variety of different colloidal particles which relative concentration decrease in the row: hydrous ferric oxides (HFO) >> clays ≅ calcite > rutile ≅ hematite ≅ barite ≅ MnO 2 > monazite > other phases. The SIMS with submicron resolution (Cameca nanoSIMS-50) was used to study local concentration of actinides. According to the obtained data among different colloids detected in the sample actinides were preferentially bound to HFO and MnO 2 while other phases did not sorb actinides. In order to determine actinide speciation bound to HFO colloids XPS and An L 3 edge XAFS measurements were done at Siberian Synchrotron Radiation Centre. The storage ring VEPP-3 with electron beam energy of 2 GeV and an average stored current of 80 mA was used as the source of radiation. Since the concentration of actinides in actual samples was too low for XAFS, the samples for measurements were prepared by contacting about 10 -5 M solutions of Np(V) and Pu(V) with

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.

Studies of glass waste form performance at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Banba, Tsunetaka; Kamizono, Hiroshi; Nakayama, Shinichi; Tashiro, Shingo

1989-08-01

The recent studies of glass waste form performance at Japan Atomic Energy Research Institute can be classified into the following three categories; (1) Study on the volatilization of radionuclides from the waste glass, which is necessary to estimate the safety in relation to operation of a storage facility. (2) Study on the radiation (alpha-radiation) effects which have relation to the long-term stability of the waste glass. (3) Study on the leaching behavior of actinides under the repository conditions, which is necessary to predict the long-term release rate of radionuclides from the waste glass. In the present report, the recent results corresponding to the above categories are described. (author)

Actinide recovery techniques utilizing electromechanical processes

International Nuclear Information System (INIS)

Westphal, B.R.; Benedict, R.W.

1994-01-01

Under certain conditions, the separation of actinides using electromechanical techniques may be an effective means of residue processing. The separation of granular mixtures of actinides and other materials discussed in this report is based on appreciable differences in the magnetic and electrical properties of the actinide elements. In addition, the high density of actinides, particularly uranium and plutonium, may render a simultaneous separation based on mutually complementary parameters. Both high intensity magnetic separation and electrostatic separation have been investigated for the concentration of an actinide waste stream. Waste stream constituents include an actinide metal alloy and broken quartz shards. The investigation of these techniques is in support of the Integral Fast Reactor (IFR) concept currently being developed at Argonne National Laboratory under the auspices of the Department of Energy

Studies of high-level waste form performance at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Banba, Tsunetaka; Mitamura, Hisayoshi; Kuramoto, Kenichi; Kamizono, Hiroshi; Inagaki, Yahohiro.

1998-02-01

The JAERI studies on the properties of the glass and ceramic waste forms, which have been done in the last several years, are described briefly. For the long-term evaluation of glass waste form performance under repository condition, leachability has studied from the standpoints of understanding of alteration layers, effects of groundwater and effects of redox condition using the radioactive or non-radioactive glass samples. The studies revealed that (1) the reactions in the alteration layers, such as crystal growth, continue after the apparent release of elements from the glass almost ceases, (2) under somewhat reducing conditions, Fe dissolves easily into leachates, and hydrated silicate surface layer tends to dissolve more easily with Fe in reduced synthetic groundwater than in deionized water, (3) precipitation of PuO 2 ·xH 2 O(am) is controlling the leaching of soluble species of Pu under both redox conditions, and the dominant soluble species is Pu(OH) 4 0 under reducing condition. Ceramics are considered as most promising materials for the actinide-rich wastes arising from partitioning and transmutation processes because of their outstanding durability for long term. In the present study, α-decay damage effects on the density and leaching behavior of perovskite (1 of 3 main minerals forming Synroc) were investigated by an accelerated experiment using the actinide doping technique. A decrease in density of Cm-doped perovskite reaches 1.3% at a dose of 9x10 17 α-decays·g -1 . The leach rate of perovskite increases with an increase in accumulated α-decay doses. Application of zirconia- and alumina-based ceramics for incorporating actinides was also investigated by inactive laboratory tests with an emphasis on crystallographic phase stability and chemical durability. The yttria-stabilized zirconia is stable crystallographically in the wide ranges of Ce and/or Nd content and have excellent chemical durability. (author)

Study of the actinide-lanthanide separation from nuclear waste by a new pyrochemical process

International Nuclear Information System (INIS)

Lemort, F.

1997-01-01

The theoretical extraction and separation of platinoids, actinides and lanthanides is allowed by thermodynamic using two adapted reducing agents: zinc and magnesium. Thereby, a pyrochemical method for the nuclear waste processing has been devised. The high temperature handling of the elements in fluoride forms and their processing by a reactive metallic phase required special precautions. The study of the behavior of matter in exploratory systems allowed the development of an experimental technology for the treatment and contacting of phases. The thermodynamical analysis of the experimental results shows the feasibility of the process. A model was developed to predict the distribution coefficients of zirconium, uranium and lanthanum as a function of the system composition. An estimation method was proposed in order to evaluate the distribution coefficients in diluted solution of all the actinides and lanthanides existing in the fission products between LiF CaF 2 and Zn-Mg at 720 deg C. Coupled with the experimental results, the estimates results may be extrapolated to concentrated solutions allowing predictions of the separation of all actinides and lanthanides. The rapidity of element transfer is induced by a thermal effect caused by the high exothermicity of the reduction by magnesium. The kinetic coefficients have been linked with the reduction enthalpy of each element. Moreover, the kinetics seem limited by chemical reaction and not by mass transfer. (author)

Criteria for achieving actinide reduction goals

International Nuclear Information System (INIS)

Liljenzin, J.O.

1996-01-01

In order to discuss various criteria for achieving actinide reduction goals, the goals for actinide reduction must be defined themselves. In this context the term actinides is interpreted to mean plutonium and the so called ''minor actinides'' neptunium, americium and curium, but also protactinium. Some possible goals and the reasons behind these will be presented. On the basis of the suggested goals it is possible to analyze various types of devices for production of nuclear energy from uranium or thorium, such as thermal or fast reactors and accelerator driven system, with their associated fuel cycles with regard to their ability to reach the actinide reduction goals. The relation between necessary single cycle burn-up values, fuel cycle processing losses and losses to waste will be defined and discussed. Finally, an attempt is made to arrange the possible systems on order of performance with regard to their potential to reduce the actinide inventory and the actinide losses to wastes. (author). 3 refs, 3 figs, 2 tabs

Plutonium-238 alpha-decay damage study of the ceramic waste form

International Nuclear Information System (INIS)

Frank, S. M.; Barber, T. L.; Cummings, D.G.; DiSanto, T.; Esh, D.W.; Giglio, J. J.; Goff, K. M.; Johnson, S.G.; Kennedy, J.R.; Jue, J-F; Noy, M.; O'Holleran, T.P.; Sinkler, W.

2006-01-01

, presumably due to alpha-decay damage. (5) No bulk sample swelling was observed. (6) No amorphization of sodalite or actinide bearing phases was observed after four years of alpha-decay damage. (7) No microcracks or phase de-bonding were observed in waste form samples aged for four years. (8) In some areas of the 238 Pu doped ceramic waste form material bubbles and voids were found. Bubbles and voids with similar size and density were also found in ceramic waste form samples without actinide. These bubbles and voids are interpreted as pre-existing defects. However, some contribution to these bubbles and voids from helium gas can not be ruled out. (9) Chemical durability of 238 Pu CWF has not changed significantly after four years of alpha-decay exposure except for an increase in the release of salt components and Pu. Still, the plutonium release from CWF is very low at less than 0.005 g/m 2

α-waste conditioning concepts on the basis of waste arisings, actinide distribution and their influence on final disposal products

International Nuclear Information System (INIS)

Krause, H.; Scheffler, K.

1978-01-01

Among the wastes arising from the reprocessing and Pu-fuel element fabrication plants, only seven waste streams contain the major part of the actinides going into the radioactive waste. It is shown that the liquid α-waste from fuel element fabrication, the high level liquid waste, and the active fraction of the medium level liquid waste can be incorporated into borosilicate glass. Wet combustion of solid burnable waste allows a relatively easy and complete recovery of plutonium. Leached hulls, sludges from feed clarification and solid non-combustible wastes can be incorporated into concrete. These treatment methods guarantee that only relatively small amounts of high quality α-bearing residues have to be disposed of

Subsurface Biogeochemistry of Actinides

Energy Technology Data Exchange (ETDEWEB)

Kersting, Annie B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Univ. Relations and Science Education; Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Glenn T. Seaborg Inst.

2016-06-29

A major scientific challenge in environmental sciences is to identify the dominant processes controlling actinide transport in the environment. It is estimated that currently, over 2200 metric tons of plutonium (Pu) have been deposited in the subsurface worldwide, a number that increases yearly with additional spent nuclear fuel (Ewing et al., 2010). Plutonium has been shown to migrate on the scale of kilometers, giving way to a critical concern that the fundamental biogeochemical processes that control its behavior in the subsurface are not well understood (Kersting et al., 1999; Novikov et al., 2006; Santschi et al., 2002). Neptunium (Np) is less prevalent in the environment; however, it is predicted to be a significant long-term dose contributor in high-level nuclear waste. Our focus on Np chemistry in this Science Plan is intended to help formulate a better understanding of Pu redox transformations in the environment and clarify the differences between the two long-lived actinides. The research approach of our Science Plan combines (1) Fundamental Mechanistic Studies that identify and quantify biogeochemical processes that control actinide behavior in solution and on solids, (2) Field Integration Studies that investigate the transport characteristics of Pu and test our conceptual understanding of actinide transport, and (3) Actinide Research Capabilities that allow us to achieve the objectives of this Scientific Focus Area (SFA and provide new opportunities for advancing actinide environmental chemistry. These three Research Thrusts form the basis of our SFA Science Program (Figure 1).

Actinide nanoparticle research

International Nuclear Information System (INIS)

Kalmykov, Stepan N.; Denecke, Melissa A.

2011-01-01

This is the first book to cover actinide nano research. It is of interest both for fundamental research into the chemistry and physics of f-block elements as well as for applied researchers such as those studying the long-term safety of nuclear waste disposal and developing remediation strategies. The authors cover important issues of the formation of actinide nano-particles, their properties and structure, environmental behavior of colloids and nanoparticles related to the safe disposal of nuclear wastes, modeling and advanced methods of characterization at the nano-scale. (orig.)

CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT

Energy Technology Data Exchange (ETDEWEB)

Brinkman, K.; Fox, K.; Marra, J.

2012-05-15

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets

Flexible waste management to increase the effectiveness of minor actinide PT technology

Energy Technology Data Exchange (ETDEWEB)

Fukasawa, T. [Hitachi-GE Nuclear Energy, Ltd., 3-1-1 Saiwai, Hitachi 317-0073 (Japan); Inagaki, Y.; Arima, T. [Kyshu University, 744 Motooka, Nishi, Fukuoka 819-0395 (Japan); Sato, S. [Fukushima National College of Technology, 30 Aza-Nagao, Tairakamiarakawa, Iwaki 970-8034 (Japan)

2016-07-01

Partitioning and transmutation (PT) technologies have been developed for minor actinides (MA) to reduce the high level waste (HLW) volume and long-term radiotoxicity. Although the MA PT can reduce the potential radiotoxicity effectively by 1-3 orders of magnitude, the actual operation of PT requires several tens of years for developing elemental technologies of nuclide separation, MA containing fuel fabrication, transmutation and their practical systematization. The high level liquid waste (HLLW) containing MA is presently vitrified immediately after spent fuel reprocessing, stored about 50 years at surface facility and will be disposed of at deep geological repository. Vitrified HLW form works as an excellent artificial barrier against nuclides release during storage and disposal. On the other hand, it is difficult to recover MA from the form. So the present waste management scheme has an issue of MA PT technology application until its deployment, which will produce much amount of vitrified HLW including long-lived MA without PT application. Thus the authors proposed the flexible waste management method to increase the effectiveness of the MA PT. The system adopts the HLLW calcination instead of the vitrification to produce granule for its dry storage of about 50 years until the MA PT technology will be applicable. The granule should be easily dissolved by the nitric acid solution to apply the typical aqueous MA partitioning technologies to be developed. This paper reports the purpose of the study, the feasibility evaluation results for the calcined granule storage and the evaluation results for the environmental burden reduction effect. (authors)

Recovery actinide values

International Nuclear Information System (INIS)

Horwitz, E.P.; Delphin, W.H.; Mason, G.W.

1979-01-01

A process is described for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of di-hexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate solution of tetramethylammonium hydrogen oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values. Uranium values remain in the extractant and may be recovered with a phosphoric acid strip. The neptunium and plutonium values are recovered from the oxalate by adding sufficient nitric acid to destroy the complexing ability of the oxalate, forming a second feed, and contacting the second feed with a second extractant of tricaprylmethylammonium nitrate in an inert diluent whereby the neptunium and plutonium values are selectively extracted. The values are recovered from the extractant with formic acid. (author)

INERT-MATRIX FUEL: ACTINIDE ''BURNING'' AND DIRECT DISPOSAL

International Nuclear Information System (INIS)

Rodney C. Ewing; Lumin Wang

2002-01-01

Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada and Japan studies have determined much improved efficiencies for burnup of actinides using inert-matrix fuels. This innovative approach also considers the properties of the inert-matrix fuel as a nuclear waste form for direct disposal after one-cycle of burn-up. Direct disposal can considerably reduce cost, processing requirements, and radiation exposure to workers

Synthesis and characterization of templated ion exchange resins for the selective complexation of actinide ions. 1998 annual progress report

International Nuclear Information System (INIS)

Murray, G.M.; Uy, O.M.

1998-01-01

'The purpose of this research is to develop polymeric extractants for the selective complexation of uranyl ions (and subsequently other actinyl and actinide ions) from aqueous solutions (lakes, streams, waste tanks and body fluids). Selectivity for a specific actinide ion is obtained by providing polymers with cavities lined with complexing ligands so arranged as to match the charge, coordination number, coordination geometry, and size of the actinide metal ion. These cavity-containing polymers will be produced using a specific actinide ion (or surrogate) as a template around which monomeric complexing ligands will be polymerized. The polymers will provide useful sequestering agents for removing actinide ions from wastes and will form the basis for a variety of analytical techniques for actinide determinations.'

Study of extraterrestrial disposal of radioactive wastes. Part 3: Preliminary feasibility screening study of space disposal of the actinide radioactive wastes with 1 percent and 0.1 percent fission product contamination

Science.gov (United States)

Hyland, R. E.; Wohl, M. L.; Finnegan, P. M.

1973-01-01

A preliminary study was conducted of the feasibility of space disposal of the actinide class of radioactive waste material. This waste was assumed to contain 1 and 0.1 percent residual fission products, since it may not be feasible to completely separate the actinides. The actinides are a small fraction of the total waste but they remain radioactive much longer than the other wastes and must be isolated from human encounter for tens of thousands of years. Results indicate that space disposal is promising but more study is required, particularly in the area of safety. The minimum cost of space transportation would increase the consumer electric utility bill by the order of 1 percent for earth escape and 3 percent for solar escape. The waste package in this phase of the study was designed for normal operating conditions only; the design of next phase of the study will include provisions for accident safety. The number of shuttle launches per year required to dispose of all U.S. generated actinide waste with 0.1 percent residual fission products varies between 3 and 15 in 1985 and between 25 and 110 by 2000. The lower values assume earth escape (solar orbit) and the higher values are for escape from the solar system.

Demonstration of innovative partitioning processes for minor actinide recycling from high active waste solutions

International Nuclear Information System (INIS)

Modolo, G.; Wilden, A.; Geist, A.; Malmbeck, R.; Taylor, R.

2014-01-01

The recycling of the minor actinides (MA) using the Partitioning and Transmutation strategy (P and T) could contribute significantly to reducing the volume of high level waste in a geological repository and to decreasing the waste's longterm hazards originating from the long half-life of the actinides. Several extraction processes have been developed worldwide for the separation and recovery of MA from highly active raffinates (HAR, e.g. the PUREX raffinate). A multi-cycle separation strategy has been developed within the framework of European collaborative projects. The multi-cycle processes, on the one hand, make use of different extractants for every single process. Within the recent FP7 European research project ACSEPT (Actinide reCycling by SEParation and Transmutation), the development of new innovative separation processes with a reduced number of cycles was envisaged. In the so-called 'innovative SANEX' concept, the trivalent actinides and lanthanides are co-extracted from the PUREX raffinate by a DIAMEX like process (e.g. TODGA). Then, the loaded solvent is subjected to several stripping steps. The first one concerns selectively stripping the actinides(III) with selective water-soluble ligands (SO3-Ph-BTB), followed by the subsequent stripping of trivalent lanthanides. A more challenging route studied also within our laboratories is the direct actinide(III) separation from a PUREX-type raffinate using a mixture of CyMe 4 BTBP and TODGA as extractants, the so-called One cycle SANEX process. A new approach, which was also studied within the ACSEPT project, is the GANEX (Grouped ActiNide EXtraction) concept addressing the simultaneous partitioning of all transuranium (TRU) elements for their homogeneous recycling in advanced generation IV reactor systems. Bulk uranium is removed in the GANEX 1st cycle, e.g. using a monoamide extractant and the GANEX 2nd cycle then separates the TRU. A solvent composed of TODGA + DMDOHEMA in kerosene has been shown to

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-01-17

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

Materials interactions test methods to measure radionuclide release from waste forms under repository-relevant conditions

International Nuclear Information System (INIS)

Strickert, R.G.; Erikson, R.L.; Shade, J.W.

1984-10-01

At the request of the Basalt Waste Isolation Project, the Materials Characterization Center has collected and developed a set of procedures into a waste form compliance test method (MCC-14.4). The purpose of the test is to measure the steady-state concentrations of specified radionuclides in solutions contacting a waste form material. The test method uses a crushed waste form and basalt material suspended in a synthetic basalt groundwater and agitated for up to three months at 150 0 C under anoxic conditions. Elemental and radioisotopic analyses are made on filtered and unfiltered aliquots of the solution. Replicate experiments are performed and simultaneous tests are conducted with an approved test material (ATM) to help ensure precise and reliable data for the actual waste form material. Various features of the test method, equipment, and test conditions are reviewed. Experimental testing using actinide-doped borosilicate glasses are also discussed. 9 references, 2 tables

Radiation damage in nuclear waste ceramics

International Nuclear Information System (INIS)

Turcotte, R.P.; Roberts, F.P.; Rusin, J.M.; Wald, J.W.

1982-01-01

The text contains a number of specific observations about the radiation-induced changes in glass, glass-ceramic, and supercalcine nuclear waste forms. Other, more general conclusions can be summarized: Radiation-induced property changes follow an exponential ingrowth curve to saturation. Actinide host phases in both crystalline waste forms become X-ray amorphous. The magnitudes of the waste-form density changes observed could not be directly related to observed changes in the primary actinide phases. Although large crystal-structure changes occur in the materials studied, obvious physical degradation was not observed

Supercritical Carbon Dioxide-Soluble Ligands for Extracting Actinide Metal Ions from Porous Solids

International Nuclear Information System (INIS)

Dietz, Mark L.

2001-01-01

Numerous types of actinide-bearing waste materials are found throughout the DOE complex. Most of these wastes consist of large volumes of non-hazardous materials contaminated with relatively small quantities of actinide elements. Separation of these wastes into their inert and radioactive components would dramatically reduce the costs of stabilization and disposal. For example, the DOE is responsible for decontaminating concrete within 7000 surplus contaminated buildings. The best technology now available for removing surface contamination from concrete involves removing the surface layer by grit blasting, which produces a large volume of blasting residue containing a small amount of radioactive material. Disposal of this residue is expensive because of its large volume and fine particulate nature. Considerable cost savings would result from separation of the radioactive constituents and stabilization of the concrete dust. Similarly, gas diffusion plants for uranium enrichment contain valuable high-purity nickel in the form of diffusion barriers. Decontamination is complicated by the extremely fine pores in these barriers, which are not readily accessible by most cleaning techniques. A cost-effective method for the removal of radioactive contaminants would release this valuable material for salvage. The objective of this project is to develop novel, substituted diphosphonic acid ligands that can be used for supercritical carbon dioxide extraction of actinide ions from solid wastes. Specifically, selected diphosphonic acids, which are known to form extremely stable complexes with actinides in aqueous and organic solution, are to be rendered carbon dioxide-soluble by the introduction of appropriate alkyl- or silicon-containing substituents. The metal complexation chemistry of these new ligands in SC-CO2 will then be investigated and techniques for their use in actinide extraction from porous solids developed

Chemical compatibility of HLW borosilicate glasses with actinides

International Nuclear Information System (INIS)

Walker, C.T.; Scheffler, K.; Riege, U.

1978-11-01

During liquid storage of HLLW the formation of actinide enriched sludges is being expected. Also during melting of HLW glasses an increase of top-to-bottom actinide concentrations can take place. Both effects have been studied. Besides, the vitrification of plutonium enriched wastes from Pu fuel element fabrication plants has been investigated with respect to an isolated vitrification process or a combined one with the HLLW. It is shown that the solidification of actinides from HLLW and actinide waste concentrates will set no principal problems. The leaching of actinides has been measured in salt brine at 23 0 C and 115 0 C. (orig.) [de

Separations chemistry for actinide elements: Recent developments and historical perspective

International Nuclear Information System (INIS)

Nash, K.L.; Choppin, G.R.

1997-01-01

With the end of the cold war, the principal mission in actinide separations has changed from production of plutonium to cleanup of the immense volume of moderately radioactive mixed wastes which resulted from fifty years of processing activities. In order to approach the cleanup task from a proper perspective, it is necessary to understand how the wastes were generated. Most of the key separations techniques central to actinide production were developed in the 40's and 50's for the identification and production of actinide elements. Total actinide recovery, lanthanide/actinide separations, and selective partitioning of actinides from inert constituents are currently of primary concern. To respond to the modern world of actinide separations, new techniques are being developed for separations ranging from analytical methods to detect ultra-trace concentrations (for bioassay and environmental monitoring) to large-scale waste treatment procedures. In this report, the history of actinide separations, both the basic science and production aspects, is examined and evaluated in terms of contemporary priorities

Ceramic waste forms for fuel-containing masses at Chernobyl

International Nuclear Information System (INIS)

Oversby, V.M.

1994-05-01

The fuel materials originally in the core of the Chernobyl Unit 4 reactor are now present within the Ukrytie in three major forms: (1) very fine particles of fuel dispersed as dust (about 10 tonnes), (2) fragments of the destroyed core, and (3) lavas containing fuel, cladding, and other materials. All of these materials will need to be immobilized into waste forms suitable for final disposal. We propose a ceramic waste form system that could accommodate all three waste types with a single set of processing equipment. The waste form would include the mineral zirconolite for immobilization of actinide materials (including uranium), perovskite, nepheline, spinel, and other phases as dictated by the chemistry of the lava masses. Waste loadings as high as 50% U can be achieved if pyrochlore, a close relative of zirconolite, is used as the U host. The ceramic immobilization could be achieved with low additions of inert chemicals to minimize the final disposal volume while ensuring a durable product. The sequence of processing would be to collect and immobilize the fuel dust first. This material will require minimal preprocessing and will provide experience in the handling of the fuel materials. Core fragments would be processed next, using a cryogenic crushing stage to reduce the size prior to adding ceramic additives. The lavas would be processed last, which is compatible with the likely sequence of availability of materials and with the complexity of the operations. The lavas will require more adjustment of chemical additive composition than the other streams to ensure that the desired phases are produced in the waste form

Lessons Learned from Characterization, Performance Assessment, and EPA Regulatory Review of the 1996 Actinide Source Term for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Larson, K.W.; Moore, R.C.; Nowak, E.J.; Papenguth, H.W.; Jow, H.

1999-01-01

The Waste Isolation Pilot Plant (WIPP) is a US Department of Energy (DOE) facility for the permanent disposal of transuranic waste from defense activities. In 1996, the DOE submitted the Title 40 CFR Part 191 Compliance Certification Application for the Waste Isolation Pilot Plant (CCA) to the US Environmental Protection Agency (EPA). The CCA included a probabilistic performance assessment (PA) conducted by Sandia National Laboratories to establish compliance with the quantitative release limits defined in 40 CFR 191.13. An experimental program to collect data relevant to the actinide source term began around 1989, which eventually supported the 1996 CCA PA actinide source term model. The actinide source term provided an estimate of mobile dissolved and colloidal Pu, Am, U, Th, and Np concentrations in their stable oxidation states, and accounted for effects of uncertainty in the chemistry of brines in waste disposal areas. The experimental program and the actinide source term included in the CCA PA underwent EPA review lasting more than 1 year. Experiments were initially conducted to develop data relevant to the wide range of potential future conditions in waste disposal areas. Interim, preliminary performance assessments and actinide source term models provided insight allowing refinement of experiments and models. Expert peer review provided additional feedback and confidence in the evolving experimental program. By 1995, the chemical database and PA predictions of WIPP performance were considered reliable enough to support the decision to add an MgO backfill to waste rooms to control chemical conditions and reduce uncertainty in actinide concentrations, especially for Pu and Am. Important lessons learned through the characterization, PA modeling, and regulatory review of the actinide source term are (1) experimental characterization and PA should evolve together, with neither activity completely dominating the other, (2) the understanding of physical processes

Actinide separations by supported liquid membranes

International Nuclear Information System (INIS)

Danesi, P.R.; Horwitz, E.P.; Rickert, P.; Chiarizia, R.

1984-01-01

The work has demonstrated that actinide removal from synthetic waste solutions using both flat-sheet and hollow-fiber SLM's is a feasible chemical process at the laboratory scale level. The process is characterized by the typical features of SLM's processes: very small quantities of extractant required; the potential for operations with high feed/strip volume ratios, resulting in a corresponding concentration factor of the actinides; and simplicity of operation. Major obstacles to the implementation of the SLM technology to the decontamination of liquid nuclear wastes are the probable low resistance of polypropylene supports to high radiation fields, which may prevent the application to high-level nuclear wastes; the unknown lifetime of the SLM; and the high Na content of the separated actinide solution

Safe management of actinides in the nuclear fuel cycle: Role of mineralogy

International Nuclear Information System (INIS)

Ewing, R.C.

2011-01-01

During the past 60 years, more than 1800 metric tonnes of Pu, and substantial quantities of the 'minor' actinides, such as Np, Am and Cm, have been generated in nuclear reactors. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239 Pu), a source of fissile material for nuclear weapons (e.g., 239 Pu and 237 Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239 Pu and 237 Np). There are two basic strategies for the disposition of these heavy elements: (1) to 'burn' or transmute the actinides using nuclear reactors or accelerators; (2) to 'sequester' the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, especially isometric pyrochlore, A 2 B 2 O 7 (A rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage. Recent developments in our understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms. (author)

Production and measurement of minor actinides in the commercial fuel cycle

International Nuclear Information System (INIS)

Stanbro, W.D.

1997-03-01

The minor actinide elements, particularly neptunium and americium, are produced as a normal byproduct of the operation of thermal power reactors. Because of the existence of long-lived isotopes of these elements, they constitute the major sources of the residual radiation in spent fuel or in wastes resulting from reprocessing. This has led to examinations by some countries of the possibility of separating the minor actinides from waste products. The papers found in this report address the production of minor actinides in common thermal power reactors as well as approaches to measure these materials in various media. The first paper in this volume, open-quotes Production of Minor Actinides in the Commercial Fuel Cycle,close quotes uses calculations with the ORIGEN2 reactor and decay code to estimate the amounts of minor actinides in spent fuel and separated plutonium as a function of reactor irradiation and the time after discharge. The second paper, open-quotes Destructive Assay of Minor Actinides,close quotes describes a number of promising approaches for the chemical analysis of minor actinides in the various forms in which they are found at reprocessing plants. The next paper, open-quotes Hybrid KED/XRF Measurement of Minor Actinides in Reprocessing Plants,close quotes uses the results of a simulation model to examine the possible applications of the hybrid KED/XRF instrument to the determination of minor actinides in some of the solutions found in reprocessing plants. In open-quotes Calorimetric Assay of Minor Actinides,close quotes the authors show some possible extensions of this powerful technique beyond the normal plutonium assays to include the minor actinides. Finally, the last paper in this volume, open-quotes Environment Measurements of Transuranic Nuclides,close quotes discusses what is known about the levels of the minor actinides in the environment and ways to analyze for these materials in environmental matrices

Actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP): FY94 results

Energy Technology Data Exchange (ETDEWEB)

Novak, C.F. [ed.

1995-08-01

This document contains six reports on actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP). These reports, completed in FY94, are relevant to the estimation of the potential dissolved actinide concentrations in WIPP brines under repository breach scenarios. Estimates of potential dissolved actinide concentrations are necessary for WIPP performance assessment calculations. The specific topics covered within this document are: the complexation of oxalate with Th(IV) and U(VI); the stability of Pu(VI) in one WIPP-specific brine environment both with and without carbonate present; the solubility of Nd(III) in a WIPP Salado brine surrogate as a function of hydrogen ion concentration; the steady-state dissolved plutonium concentrations in a synthetic WIPP Culebra brine surrogate; the development of a model for Nd(III) solubility and speciation in dilute to concentrated sodium carbonate and sodium bicarbonate solutions; and the development of a model for Np(V) solubility and speciation in dilute to concentrated sodium Perchlorate, sodium carbonate, and sodium chloride media.

Actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP): FY94 results

International Nuclear Information System (INIS)

Novak, C.F.

1995-08-01

This document contains six reports on actinide chemistry research supporting the Waste Isolation Pilot Plant (WIPP). These reports, completed in FY94, are relevant to the estimation of the potential dissolved actinide concentrations in WIPP brines under repository breach scenarios. Estimates of potential dissolved actinide concentrations are necessary for WIPP performance assessment calculations. The specific topics covered within this document are: the complexation of oxalate with Th(IV) and U(VI); the stability of Pu(VI) in one WIPP-specific brine environment both with and without carbonate present; the solubility of Nd(III) in a WIPP Salado brine surrogate as a function of hydrogen ion concentration; the steady-state dissolved plutonium concentrations in a synthetic WIPP Culebra brine surrogate; the development of a model for Nd(III) solubility and speciation in dilute to concentrated sodium carbonate and sodium bicarbonate solutions; and the development of a model for Np(V) solubility and speciation in dilute to concentrated sodium Perchlorate, sodium carbonate, and sodium chloride media

Use of natural analogues to predict the behavior of transuranic actinide elements in the environment

International Nuclear Information System (INIS)

Eisenbud, M.; Linsalata, P.; Penna Franca, E.

1986-01-01

An important question that must be considered in assessing the long-term effect of high-level waste isolation in geological repositories is how the actinide elements would behave if a nuclear waste repository should be exposed by erosion or be intruded by ground water. A substantial literature has accumulated concerning the behavior of the transuranic actinide elements in the environment, but most of the research until now has been concerned with fallout from nuclear weapons tests or with contamination in the environments of major atomic energy production plants. These studies are only partially relevant to risk assessment of stored radioactive wastes because such contaminants have chemical and physical forms different from those that would exist in a nuclear waste repository. Moreover, the long-term effects of geochemical processes on the behavior of the actinide elements cannot be studied because these elements were first produced by artificial means only 40 years ago

Storage of plutonium and nuclear power plant actinide waste in the form of critical-mass-free ceramics containing neutron poisons

Energy Technology Data Exchange (ETDEWEB)

Nadykto, B.A. [RFNC-VNIIEF, Nizhni Novgorod Region (Russian Federation)

2001-07-01

The nuclear weapons production has resulted in accumulation of a large quantity of plutonium and uranium highly enriched with uranium-235 isotope (many tons). The work under ISTC Project 332B-97 treated the issues of safe plutonium storage through making critical-mass-free plutonium oxide compositions with neutron poisons. This completely excludes immediate utilization (without chemical reprocessing) of retained plutonium in nuclear devices. It is therewith possible to locate plutonium most compactly in the storage facility, which would allow reduction in required storage areas and costs. The issues of the surplus weapon-grade plutonium management and utilization have been comprehensively studied in the recent decade. The issues are treated in multiple scientific publications, conferences, and seminars. At the same time, issues of nuclear power engineering actinide waste storage are studied no less extensively. The general issues are material radioactivity and energy release and nuclear accident hazards due to critical mass generation. Plutonium accumulated in nuclear power plant spent fuel is more accessible than weapon-grade plutonium and can become of higher and higher interest with time as its activity reduces, including as material for nuclear devices. The urgency of plutonium management is presently related not only to accumulation of surplus weapon-grade plutonium, but also to the fact that it is high time to decide what has to be done regarding reactor plutonium. Presently, the possibility of actinide separation from NPP spent nuclear fuel and compact underground burial separately from other (mainly fragment) activity is being considered. Actinide and neutron poison base critical-mass-free ceramic materials (similar to plutonium ceramics) may be useful for this burial method. (author)

Storage of plutonium and nuclear power plant actinide waste in the form of critical-mass-free ceramics containing neutron poisons

International Nuclear Information System (INIS)

Nadykto, B.A.

2001-01-01

The nuclear weapons production has resulted in accumulation of a large quantity of plutonium and uranium highly enriched with uranium-235 isotope (many tons). The work under ISTC Project 332B-97 treated the issues of safe plutonium storage through making critical-mass-free plutonium oxide compositions with neutron poisons. This completely excludes immediate utilization (without chemical reprocessing) of retained plutonium in nuclear devices. It is therewith possible to locate plutonium most compactly in the storage facility, which would allow reduction in required storage areas and costs. The issues of the surplus weapon-grade plutonium management and utilization have been comprehensively studied in the recent decade. The issues are treated in multiple scientific publications, conferences, and seminars. At the same time, issues of nuclear power engineering actinide waste storage are studied no less extensively. The general issues are material radioactivity and energy release and nuclear accident hazards due to critical mass generation. Plutonium accumulated in nuclear power plant spent fuel is more accessible than weapon-grade plutonium and can become of higher and higher interest with time as its activity reduces, including as material for nuclear devices. The urgency of plutonium management is presently related not only to accumulation of surplus weapon-grade plutonium, but also to the fact that it is high time to decide what has to be done regarding reactor plutonium. Presently, the possibility of actinide separation from NPP spent nuclear fuel and compact underground burial separately from other (mainly fragment) activity is being considered. Actinide and neutron poison base critical-mass-free ceramic materials (similar to plutonium ceramics) may be useful for this burial method. (author)

Research for actinides extractants from various wastes

International Nuclear Information System (INIS)

Musikas, C.; Cuillerdier, C.; Condamines, N.

1990-01-01

This paper is an overview of the actinides solvent extraction research undertaken in Fontenay-aux-Roses. Two kinds of extractants are investigated; those usable for the improvement of the nowadays nuclear fuels reprocessing and those necessary for advanced fuels cycles which include the minor actinides (Np, Am) recovery for a further elimination through nuclear reactions. In the first class the mono and diamides, alternative to the organophosphorus extractants, TBP and polyfunctional phosphonates, showed promising properties. The main results are discussed. For the future efficient extractants for trivalent actinides-lanthanides group separations are suitable. The point about the actinides (III) - lanthanides (III) group separation chemistry and the development of some of these extractants are given

Studies of high-level waste form performance at Japan Atomic Energy Research Institute

Energy Technology Data Exchange (ETDEWEB)

Banba, Tsunetaka; Mitamura, Hisayoshi; Kuramoto, Kenichi; Kamizono, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Inagaki, Yahohiro

1998-02-01

The JAERI studies on the properties of the glass and ceramic waste forms, which have been done in the last several years, are described briefly. For the long-term evaluation of glass waste form performance under repository condition, leachability has studied from the standpoints of understanding of alteration layers, effects of groundwater and effects of redox condition using the radioactive or non-radioactive glass samples. The studies revealed that (1) the reactions in the alteration layers, such as crystal growth, continue after the apparent release of elements from the glass almost ceases, (2) under somewhat reducing conditions, Fe dissolves easily into leachates, and hydrated silicate surface layer tends to dissolve more easily with Fe in reduced synthetic groundwater than in deionized water, (3) precipitation of PuO{sub 2}{center_dot}xH{sub 2}O(am) is controlling the leaching of soluble species of Pu under both redox conditions, and the dominant soluble species is Pu(OH){sub 4}{sup 0} under reducing condition. Ceramics are considered as most promising materials for the actinide-rich wastes arising from partitioning and transmutation processes because of their outstanding durability for long term. In the present study, {alpha}-decay damage effects on the density and leaching behavior of perovskite (1 of 3 main minerals forming Synroc) were investigated by an accelerated experiment using the actinide doping technique. A decrease in density of Cm-doped perovskite reaches 1.3% at a dose of 9x10{sup 17} {alpha}-decays{center_dot}g{sup -1}. The leach rate of perovskite increases with an increase in accumulated {alpha}-decay doses. Application of zirconia- and alumina-based ceramics for incorporating actinides was also investigated by inactive laboratory tests with an emphasis on crystallographic phase stability and chemical durability. The yttria-stabilized zirconia is stable crystallographically in the wide ranges of Ce and/or Nd content and have excellent

Transmutation of waste actinides in light water reactors

International Nuclear Information System (INIS)

Gorrell, T.C.

1979-04-01

Actinide recycle and transmutation calculations were made for three irradiation options of a light water reactor (LWR). The cases considered were: all actinides recycled in regular uranium fuel assemblies; transuranic actinides recycled in separate MOX assemblies with 235 U enrichment of uranium; and transuranic actinides recycled in separate MOX assemblies with plutonium enrichment of natural uranium. When all actinides were recycled in a uniform lattice, the transuranic inventory after ten recycles was 38% of the inventory accumulated without recycle. When the transuranics from two regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after five recycles

Actinide recycle potential in the integral fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.

1993-01-01

The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. In the IFR pyroprocessing, minor actinides accompany plutonium product stream, and therefore, actinide recycle occurs naturally. The fast neutron spectrum of the IFR makes it an ideal actinide burner, as well. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and potential implications on long-term waste management

Radiation damage in natural materials: implications for radioactive waste forms

International Nuclear Information System (INIS)

Ewing, R.C.

1981-01-01

The long-term effect of radiation damage on waste forms, either crystalline or glass, is a factor in the evaluation of the integrity of waste disposal mediums. Natural analogs, such as metamict minerals, provide one approach for the evaluaton of radiation damage effects that might be observed in crystalline waste forms, such as supercalcine or synroc. Metamict minerals are a special class of amorphous materials which were initially crystalline. Although the mechanism for the loss of crystallinity in these minerals (mostly actinide-containing oxides and silicates) is not clearly understood, damage caused by alpha particles and recoil nuclei is critical to the metamictization process. The study of metamict minerals allows the evaluation of long-term radiation damage effects, particularly changes in physical and chemical properties such as microfracturing, hydrothermal alteration, and solubility. In addition, structures susceptible to metamictization share some common properties: (1) complex compositions; (2) some degree of covalent bonding, instead of being ionic close-packed MO/sub x/ structures; and (3) channels or interstitial voids which may accommodate displaced atoms or absorbed water. On the basis of these empirical criteria, minerals such as pollucite, sodalite, nepheline and leucite warrant careful scrutiny as potential waste form phases. Phases with the monazite or fluorite structures are excellent candidates

The effects of actinide separation on the radiological consequences of geologic disposal of high-level waste

International Nuclear Information System (INIS)

Hill, M.D.; White, I.F.; Fleishman, A.B.

1980-01-01

It has often been suggested that the potential hazard to man from the disposal of high-level radioactive waste could be reduced by removing a substantial fraction of the actinide elements. In this report the effects of actinide separation on the radiological consequences of one of the disposal options currently under consideration, that of burial in deep geologic formations, are examined. The results show that the potential radiological impact of geologic disposal of high-level waste arises from both long-lived fission products and actinides (and their daughter radionuclides). Neither class of radionuclides is of overriding importance and actinide separation would therefore reduce the radiological impact to only a limited extent and over limited periods. There might be a case for attempting to reduce doses from 237 Np. To achieve this it appears to be necessary to separate both neptunium and its precursor element americium. However, there are major uncertainties in the data needed to predict doses from 237 Np; further research is required to resolve these uncertainties. In addition, consideration should be given to alternative methods of reducing the radiological impact of geologic disposal. The conclusions of this assessment differ considerably from those of similar studies based on the concept of toxicity indices. Use of these indices can lead to incorrect allocation of research and development effort. (author)

Safe management of actinides in the nuclear fuel cycle: Role of mineralogy; La gestion des actinides dans le cycle du combustible nucleaire: le role de la mineralogie

Energy Technology Data Exchange (ETDEWEB)

Ewing, R.C. [Department of Nuclear Engineering and Radiological Sciences, Department of Geological Sciences, Department of Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan 48109-1005 (United States)

2011-02-15

During the past 60 years, more than 1800 metric tonnes of Pu, and substantial quantities of the 'minor' actinides, such as Np, Am and Cm, have been generated in nuclear reactors. Some of these transuranium elements can be a source of energy in fission reactions (e.g., {sup 239}Pu), a source of fissile material for nuclear weapons (e.g., {sup 239}Pu and {sup 237}Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., {sup 239}Pu and {sup 237}Np). There are two basic strategies for the disposition of these heavy elements: (1) to 'burn' or transmute the actinides using nuclear reactors or accelerators; (2) to 'sequester' the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, especially isometric pyrochlore, A{sub 2}B{sub 2}O{sub 7} (A rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage. Recent developments in our understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms. (author)

Research on the chemical speciation of actinides

International Nuclear Information System (INIS)

Jung, Euo Chang; Park, K. K.; Cho, H. R.

2010-04-01

A demand for the safe and effective management of spent nuclear fuel and radioactive waste generated from nuclear power plant draws increasing attention with the growth of nuclear power industry. The objective of this project is to establish the basis of research on the actinide chemistry by using advanced laser-based highly sensitive spectroscopic systems. Researches on the chemical speciation of actinides are prerequisite for the development of technologies related to nuclear fuel cycles, especially, such as the safe management of high level radioactive wastes and the chemical examination of irradiated nuclear fuels. For supporting these technologies, laser-based spectroscopies have been performed for the chemical speciation of actinide in an aqueous solutions and the quantitative analysis of actinide isotopes in spent nuclear fuels. In this report, results on the following subjects have been summarized. (1) Development of TRLFS technology for chemical speciation of actinides, (2) Development of LIBD technology for measuring solubility of actinides, (3) Chemical speciation of plutonium complexes by using a LWCC system, (4) Development of LIBS technology for the quantitative analysis of actinides, (5) Development of technology for the chemical speciation of actinides by CE, (6) Evaluation on the chemical reactions between actinides and humic substances, (7) Chemical speciation of actinides adsorbed on metal oxides surfaces, (8) Determination of actinide source terms of spent nuclear fuel

Role of bacteria as biocolloids in the transport of actinides from a deep underground radioactive waste repository

International Nuclear Information System (INIS)

Francis, A.J.; Gillow, J.B.; Dodge, C.J.; Dunn, M.; Mantione, K.; Strietelmeier, B.A.; Pansoy-Hjelvik, M.E.; Papenguth, H.W.

1998-01-01

We investigated the interaction of dissolved actinides 232 Th, 238 U, 237 Np, 239 Pu, and 243 Am, with a pure and a mixed culture of halophilic bacteria isolated from the waste isolation pilot plant repository under anaerobic conditions to evaluate their potential transport as biocolloids from the waste site. The sizes of the bacterial cells studied ranged from 0.54 x 0.48 μm to 7.7 x 0.67 μm (1 x w). Using sequential microfiltration, we determined the association of actinides with free-living (mobile) bacterial cells suspended in a fluid medium containing NaCl or MgCl 2 brine, at various phases of their growth cycles. The number of suspended bacteria ranged from 10 6 to 10 9 cells ml -1 . The amount of actinide associated with the suspended cell fraction (calculated as mol cell -1 ) was very low: Th, 10 -12 ; U, 10 -15 -10 -18 ; Np, 10 -15 -10 -19 ; Pu, 10 -18 -10 -21 ; and Am, 10 -18 -10 -19 ; and it varied with the bacterial culture studied. The differences in the association are attributed to the extent of bioaccumulation and biosorption by the bacteria, pH, the composition of the brine, and the speciation and bioavailability of the actinides. (orig.)

Actinide separation chemistry in nuclear waste streams, an OECD Nuclear Energy Agency review

International Nuclear Information System (INIS)

Madic, C.

1997-01-01

The separation of actinide elements from various waste materials, either produced in nuclear fuel cycle or in the past during nuclear weapon production, represent a significant issue facing developed countries. Improvements in the efficiencies of the separation processes can be expected from a better knowledge of the chemistry of these elements in these complex matrices. The Nuclear Science Committee of the OECD/NEA has established a task force of experts in actinide chemistry to review the current and developing separation techniques and chemical processes. Recommendations were made for future research and development programs. This article presents briefly the work of the Task Force which will be published soon as an OECD/NEA/NSC Report. (author)

Investigation of microscopic radiation damage in waste forms using ODNMR and AEM techniques. 1998 annual progress report

International Nuclear Information System (INIS)

Liu, G.

1998-01-01

'This project seeks to understand the microscopic effects of radiation damage in nuclear waste forms. The authors approach to this challenge encompasses studies of crystals and glass containing short-lived alpha- and beta-emitting actinides with electron microscopy, laser spectroscopy, and computational modeling and simulation. Much of the initial effort has focused on alpha-decay induced microscopic damage in 17-year old samples of crystalline yttrium and lutetium orthophosphates and thorium dioxide that initially contained ∼1% of the alpha-emitting isotope Cm-244 (18.1 y half life) or the beta-emitting isotope Bk-249 (0.88 y half life). Studies will also be conducted on borosilicate glasses that contain Cm-244 or Am-241, respectively. The goal is to gain clear insight into accumulated radiation damage and the influence of aging on such damage, which are critical factors in the long-term performance of high-level nuclear waste forms. Amorphization previously has been thought to be the most important effect of radiation damage in crystalline and ceramic materials. The studies show that for alpha-emitting actinide ions in certain crystalline phosphates, amorphization is not a significant effect of radiation damage. Instead, formation of microscopic cavities (bubbles) is an important consequence of alpha-decay events. This amorphization-resistant property makes orthophosphates a very attractive high level nuclear waste form. However, aggregation and mobilization of cavities (bubbles) might increase the leach rate of radionuclides and influence the long-term stability of the waste forms. Further research is needed before the authors can draw a final conclusion on the long-term effects of radiation damage in high level waste forms.'

Limitations of actinide recycle and waste disposal consequences

International Nuclear Information System (INIS)

Baetsle, L.H.; Raedt, C. de

1994-01-01

The paper emphasizes the impact of Light Water Reactor - Mixed Oxides introduction on the subsequent actinide management and fate of reprocessed and depleted uranium. The spent fuel from LWR-MOX contains in principle 75% of the initially produced plutonium. This new source term has to be considered together with the minor actinides from the conventional reprocessing. Subsequent LWR-MOX reprocessing in the first step in a very long term Pu + minor actinides management. Recycling of Pu + minor actinides in fast reactors to significantly reduce the Pu and minor actinides inventory (e.g. a factor of 10) is a very slow process which requires the development and operation of a large park of actinide burner reactors during an extended period of time. The overall feasibility of the P and T option will greatly depend on the massive introduction during the next century of fast neutron reactors as a replacement to the present LWR generation of nuclear power plants. (authors). 11 refs., 6 tabs., 2 figs

Amount, disposal and relative toxicity of long-lived fission products and actinides in the radioactive wastes of the nuclear fuel cycles

International Nuclear Information System (INIS)

Haug, H.O.

1975-11-01

A review is presented on the magnitude of the long-term problems of radioactive wastes from the nuclear power industry of the FRG (and Western Europe). The production of long-lived fission products and actinides has been calculated for several fuel types of the uranium-plutonium and thorium-uranium fuel cycles and related to a prediction of the development and share of LWR, FBR and HTGR. The quantities and concentrations of actinides, the radioactivity and relative toxicity index of the wastes of reprocessing (and fuel refabrication) and their changes by radioactive decay are presented. The radiotoxicity of the nuclide inventory of the solidified high-level wastes have been compared with naturally occuring uranium ores. On the long term (>10 3 years) the radiotoxicity level of the total area of the final repository in deep geological formation does not result in a significantly higher radiotoxicity level than an uranium ore deposit of low uranium content. Also discussed have been the chemical separation of the actinides from high-level wastes and recycling in fission reactors. (orig.) [de

Research in actinide chemistry

International Nuclear Information System (INIS)

Choppin, G.R.

1993-01-01

This research studies the behavior of the actinide elements in aqueous solution. The high radioactivity of the transuranium actinides limits the concentrations which can be studied and, consequently, limits the experimental techniques. However, oxidation state analogs (trivalent lanthanides, tetravalent thorium, and hexavalent uranium) do not suffer from these limitations. Behavior of actinides in the environment are a major USDOE concern, whether in connection with long-term releases from a repository, releases from stored defense wastes or accidental releases in reprocessing, etc. Principal goal of our research was expand the thermodynamic data base on complexation of actinides by natural ligands (e.g., OH - , CO 3 2- , PO 4 3- , humates). The research undertakes fundamental studies of actinide complexes which can increase understanding of the environmental behavior of these elements

Study of the actinide-lanthanide separation from nuclear waste by a new pyrochemical process; Etude de la separation actinides-lanthanides des dechets nucleaires par un procede pyrochimique nouveau

Energy Technology Data Exchange (ETDEWEB)

Lemort, F. [CEA Marcoule, Departement de Retraitement, des Dechets et du Demantelement, 30 - Bagnols-sur-Ceze (France)]|[Institut National Polytechnique, 38 - Grenoble (France)

1997-01-01

The theoretical extraction and separation of platinoids, actinides and lanthanides is allowed by thermodynamic using two adapted reducing agents: zinc and magnesium. Thereby, a pyrochemical method for the nuclear waste processing has been devised. The high temperature handling of the elements in fluoride forms and their processing by a reactive metallic phase required special precautions. The study of the behavior of matter in exploratory systems allowed the development of an experimental technology for the treatment and contacting of phases. The thermodynamical analysis of the experimental results shows the feasibility of the process. A model was developed to predict the distribution coefficients of zirconium, uranium and lanthanum as a function of the system composition. An estimation method was proposed in order to evaluate the distribution coefficients in diluted solution of all the actinides and lanthanides existing in the fission products between LiF CaF{sub 2} and Zn-Mg at 720 deg C. Coupled with the experimental results, the estimates results may be extrapolated to concentrated solutions allowing predictions of the separation of all actinides and lanthanides. The rapidity of element transfer is induced by a thermal effect caused by the high exothermicity of the reduction by magnesium. The kinetic coefficients have been linked with the reduction enthalpy of each element. Moreover, the kinetics seem limited by chemical reaction and not by mass transfer. (author) 66 refs.

Formation of actinides in irradiated HTGR fuel elements

Energy Technology Data Exchange (ETDEWEB)

dos Santos, A. M.

1976-03-15

Actinide nuclide concentrations of 11 spent AVR fuel elements were determined experimentally. The burnup of the spheres varied in the range between 10% and 100% fifa, the Th : U ratio was 5 : 1. The separation procedures for an actinide isolation were tested with highly irradiated ThO/sub 2/. Separation and decontamination factors are presented. Build-up of /sup 232/U was discussed. The AVR breeding rate was ascertained to be 0.5. The hazard potential of high activity waste was calculated. Actinide recovery factors were proposed in order to reduce the hazard potential of the waste by an actinide removal under consideration of the reprocessing technology which is available presently.

Quantum-chemical consideration of extermal valent forms of actinides

International Nuclear Information System (INIS)

Ionova, G.V.; Pershina, V.G.; Spitsyn, V.I.

1982-01-01

Stability of valent forms of actinides that has not yet studied experimentally, is considered within the framework of quantum-chemical considerations. Oxidizing potentials E 0 for actinide elements are determined theoretically. A dependence of the definite valent state stability on relativistic effect is shown. A conclusion is made that oxidizing potential E 0 (4-5) for americium should be higher than E 0 (4-5) for plutonium. A relatively small oxidizing potential E 0 (4-5) for curium speaks about principle possibility of production of five-valent curium in solution, though it is less stable than the six-valent one. Oxidizing potential corresponding to transition of three-valent californium into the four-valent state should be less than the value adopted in literature. A relatively small oxidizing potential of californium E 0 (4-5) speaks about possible existence of five-valent californium in solution

Preliminary investigation of actinide and xenon reactivity effects in accelerator transmutation of waste high-flux systems

International Nuclear Information System (INIS)

Olson, K.R.; Henderson, D.L.

1995-01-01

The possibility of an unstable positive reactivity growth in an accelerator transmutation of waste (ATW)-type high-flux system is investigated. While it has always been clear that xenon is an important actor in the reactivity response of a system to flux changes, it has been suggested that in very high thermal flux transuranic burning systems, a positive, unstable reactivity growth could be caused by the actinides alone. Initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately. The maximum change in reactivity after a flux change caused by the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or startup. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response caused by the actinides. The capabilities and applications of both the current actinide model and the xenon model are discussed. Finally, the need for a complete dynamic model for the high-flux fluid-fueled ATW system is addressed

An investigation into the use of a fast breeder reactor to incinerate actinide waste from the U.K. nuclear power programme

International Nuclear Information System (INIS)

Harte, G.A.; Clarke, R.H.

1976-12-01

Preliminary investigations are described into the effect of actinide removal and incineration (i.e. re-irradiation in a high neutron flux) on the long term hazards of reprocessing wastes from a Magnox reactor and from a notional CFR operated on plutonium from a Magnox reactor and also from an AGR. The actinides neptunium, americium and curium extracted from these wastes would be amenable to incineration in a fast breeder reactor core with fluxes of the order of 10 16 ncm -2 s -1 . An incineration period of 35 years was found to achieve a reduction in hazard, for most cooling periods up to one million years. The reductions were 2 to 3 orders of magnitude for the Magnox wastes and 4 orders of magnitude for the waste from CFR fuelled with AGR-plutonium. The large reduction factors do not apply to the hazards of reprocessing wastes as a whole, however, since one actinide separation has been achieved the residual quantities of uranium and plutonium in the waste stream, assumed here to be 0.1% of the original fuel inventory of these elements, become dominant in determining the hazard. Shorter incineration periods may serve to reduce the hazards of the extracted transuranics below the levels determined by these residues. A discussion of the hazards associated with reprocessing wastes as compared to those associated with uranium mill tailings attempts to place the nuclear waste disposal problem in perspective. (author)

Research on the chemical speciation of actinides

International Nuclear Information System (INIS)

Jung, Euo Chang; Park, K. K.; Cho, H. R.

2012-04-01

A demand for the safe and effective management of spent nuclear fuel and radioactive waste generated from nuclear power plant draws increasing attention with the growth of nuclear power industry. The objective of this project is to establish the basis of research on the actinide chemistry by using highly sensitive and advanced laser-based spectroscopic systems. Researches on the chemical speciation of actinides are prerequisite for the development of technologies related to nuclear fuel cycles, especially, such as the safe management of high level radioactive wastes and the chemical examination of irradiated nuclear fuels. For supporting these technologies, laser-based spectroscopies have been applied for the chemical speciation of actinide in aqueous solutions and the quantitative analysis of actinide isotopes in spent nuclear fuels. In this report, results on the following subjects have been summarized. Development of TRLFS technology for the chemical speciation of actinides, Development of laser-induced photo-acoustic spectroscopy (LPAS) system, Application of LIBD technology to investigate dynamic behaviors of actinides dissolution reactions, Development of nanoparticle analysis technology in groundwater using LIBD, Chemical speciation of plutonium complexes by using a LWCC system, Development of LIBS technology for the quantitative analysis of actinides, Evaluation on the chemical reactions between actinides and humic substances, Spectroscopic speciation of uranium-ligand complexes in aqueous solution, Chemical speciation of actinides adsorbed on metal oxides surfaces

Actinide metal processing

International Nuclear Information System (INIS)

Sauer, N.N.; Watkin, J.G.

1992-01-01

A process for converting an actinide metal such as thorium, uranium, or plutonium to an actinide oxide material by admixing the actinide metal in an aqueous medium with a hypochlorite as an oxidizing agent for sufficient time to form the actinide oxide material and recovering the actinide oxide material is described together with a low temperature process for preparing an actinide oxide nitrate such as uranyl nitrate. Additionally, a composition of matter comprising the reaction product of uranium metal and sodium hypochlorite is provided, the reaction product being an essentially insoluble uranium oxide material suitable for disposal or long term storage

Characterization of actinide-bearing sediments underlying liquid waste disposal facilities at Hanford

International Nuclear Information System (INIS)

Price, S.M.; Ames, L.L.

1975-09-01

Past liquid waste disposal practices at the U. S. Energy Research and Development Administration's Hanford Reservation have included the discharges of solutions containing trace quantities of actinides directly into the ground via structures collectively termed ''trenches''. Characterization of samples from two of these trenches, the 216-Z-9 and the 216-Z-1A(a), has been initiated to determine the present form and migration potential of plutonium stored in sediments which received high salt, acidic waste liquids. Analysis of samples acquired by drilling has revealed that the greatest measured concentration of Pu, approximately 10 6 μCi 239 Pu/liter of sediment, occurs in both facilities just below the points of release of the waste liquids. This concentration decreases to approximately 10 3 μCi 239 Pu/liter of sediment within the first 2 meters of the underlying sediment columns and to approximately 10 μCi 239 Pu/liter of sediment at the maximum depth sampled (9 meters). Examination of relatively undisturbed sediment cores illustrated two types of Pu occurrence responsible for this distribution. One of these types is composed of Pu particles (greater than 70 wt percent PuO 2 ) added to the disposal site in the same form. This ''particulate'' type was ''filtered out'' within the upper 1 meter of the sediment column, accounting for the high concentration of Pu/liter of sediment in this region. The second type of Pu (less than 0.5 wt percent PuO 2 ) was originally disposed of as soluble Pu(IV). This ''nonparticulate'' type penetrated deeper within the sediment profile and was deposited in association with silicate hydrolysis of the sediment fragments

The reprocessing-recycling of spent nuclear fuel. Actinides separation - Application to wastes management; Le traitement-recyclage du combustible nucleaire use. La separation des actinides - Application a la gestion des dechets

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-07-01

After its use in the reactor, the spent fuel still contains lot of recoverable material for an energetic use (uranium, plutonium), but also fission products and minor actinides which represent the residues of nuclear reactions. The reprocessing-recycling of the spent fuel, as it is performed in France, implies the chemical separation of these materials. The development and the industrial implementation of this separation process represent a major contribution of the French science and technology. The reprocessing-recycling allows a good management of nuclear wastes and a significant saving of fissile materials. With the recent spectacular rise of uranium prices, this process will become indispensable with the development of the next generation of fast neutron reactors. This book takes stock of the present and future variants of the chemical process used for the reprocessing of spent fuels. It describes the researches in progress and presents the stakes and recent results obtained by the CEA. content: the separation of actinides, a key factor for a sustainable nuclear energy; the actinides, a discovery of the 20. century; the radionuclides in nuclear fuels; the aquo ions of actinides; some redox properties of actinides; some complexing properties of actinide cations; general considerations about treatment processes; some characteristics of nuclear fuels in relation with their reprocessing; technical goals and specific constraints of the PUREX process; front-end operations of the PUREX process; separation and purification operations of the PUREX process; elaboration of finite products in the framework of the PUREX process; management and treatment of liquid effluents; solid wastes of the PUREX process; towards a joint management of uranium and plutonium: the COEX{sup TM} process; technical options of treatment and recycling techniques; the fuels of generation IV reactors; front-end treatment processes of advanced fuels; hydrometallurgical processes for future fuel

Partitioning of actinides from high level waste of PUREX origin using octylphenyl-N,N'-diisobutylcarbamoylmethyl phosphine oxide (CMPO)-based supported liquid membrane

International Nuclear Information System (INIS)

Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Dudwadkar, N.L.; Chitnis, R.R.; Mathur, J.N.

1999-01-01

The present studies deal with the application of the supported liquid membrane (SLM) technique for partitioning of actinides from high level waste of PUREX origin. The process uses a solution of octylphenyl-N,N'-diisobutylcarbamoylmethyl phosphine oxide (CMPO) in n-dodecane as a carrier with a polytetrafluoroethylene support and a mixture of citric acid, formic acid, and hydrazine hydrate as the receiving phase. The studies involve the investigation of such parameters as carrier concentration in SLM, acidity of the feed, and the feed composition. The studies indicated good transport of actinides like neptunium, americium, and plutonium across the membrane from nitric acid medium. A high concentration of uranium in the feed retards the transport of americium, suggesting the need for prior removal of uranium from the waste. The separation of actinides from uranium-lean simulated samples as well as actual high level waste has been found to be feasible using the above technique

Use of fast reactors for actinide transmutation

International Nuclear Information System (INIS)

1993-03-01

The management of radioactive waste is one of the key issues in today's discussions on nuclear energy, especially the long term disposal of high level radioactive wastes. The recycling of plutonium in liquid metal fast breeder reactors (LMFBRs) would allow 'burning' of the associated extremely long life transuranic waste, particularly actinides, thus reducing the required isolation time for high level waste from tens of thousands of years to hundreds of years for fission products only. The International Working Group on Fast Reactors (IWGFR) decided to include the topic of actinide transmutation in liquid metal fast breeder reactors in its programme. The IAEA organized the Specialists Meeting on Use of Fast Breeder Reactors for Actinide Transmutation in Obninsk, Russian Federation, from 22 to 24 September 1992. The specialists agree that future progress in solving transmutation problems could be achieved by improvements in: Radiochemical partitioning and extraction of the actinides from the spent fuel (at least 98% for Np and Cm and 99.9% for Pu and Am isotopes); technological research and development on the design, fabrication and irradiation of the minor actinides (MAs) containing fuels; nuclear constants measurement and evaluation (selective cross-sections, fission fragments yields, delayed neutron parameters) especially for MA burners; demonstration of the feasibility of the safe and economic MA burner cores; knowledge of the impact of maximum tolerable amount of rare earths in americium containing fuels. Refs, figs and tabs

Citrate based ''TALSPEAK'' lanthanide-actinide separation process

International Nuclear Information System (INIS)

Del Cul, G.D.; Bond, W.D.; Toth, L.M.; Davis, G.D.; Dai, S.; Metcalf, D.H.

1994-09-01

The potential hazard posed to future generations by long-lived radionuclides such as the transuranic elements (TRU) is perceived as a major problem associated with the use of nuclear power. TRU wastes have to remain isolated from the environment for ''geological'' periods of time. The costs of building, maintaining, and operating a ''geological TRU repository'' can be very high. Therefore, there are significant economical advantages in segregating the relatively low volume of TRU wastes from other nuclear wastes. The chemical behavior of lanthanides and actinides, 4f and 5f elements respectively, is rather similar. As a consequence, the separation of these two groups is difficult. The ''TALSPEAK'' process (Trivalent Actinide Lanthanide Separations by Phosphorus-reagent Extraction from Aqueous Complexes) is one of the few means available to separate the trivalent actinides from the lanthanides. The method is based on the preferential complexation of the trivalent actinides by an aminopolyacetic acid. Cold experiments showed that by using citric acid the deleterious effects produced by impurities such as zirconium are greatly reduced

Development and characterization of solidified forms for high-level wastes: 1978. Annual report

International Nuclear Information System (INIS)

Ross, W.A.; Mendel, J.E.

1979-12-01

Development and characterization of solidified high-level waste forms are directed at determining both process properties and long-term behaviors of various solidified high-level waste forms in aqueous, thermal, and radiation environments. Waste glass properties measured as a function of composition were melt viscosity, melt electrical conductivity, devitrification, and chemical durability. The alkali metals were found to have the greatest effect upon glass properties. Titanium caused a slight decrease in viscosity and a significant increase in chemical durability in acidic solutions (pH-4). Aluminum, nickel and iron were all found to increase the formation of nickel-ferrite spinel crystals in the glass. Four multibarrier advanced waste forms were produced on a one-liter scale with simulated waste and characterized. Glass marbles encapsulated in a vacuum-cast lead alloy provided improved inertness with a minimal increase in technological complexity. Supercalcine spheres exhibited excellent inertness when coated with pyrolytic carbon and alumina and put in a metal matrix, but the processing requirements are quite complex. Tests on simulated and actual high-level waste glasses continue to suggest that thermal devitrification has a relatively small effect upon mechanical and chemical durabilities. Tests on the effects radiation has upon waste forms also continue to show changes to be relatively insignificant. Effects caused by decay of actinides can be estimated to saturate at near 10 19 alpha-events/cm 3 in homogeneous solids. Actually, in solidified waste forms the effects are usually observed around certain crystals as radiation causes amorphization and swelling of th crystals

Actinide separation of high-level waste using solvent extractants on magnetic microparticles

International Nuclear Information System (INIS)

Nunez, L.; Buchholz, B.A.; Kaminski, M.; Aase, S.B.; Brown, N.R.; Vandegrift, G.F.

1994-01-01

Polymeric-coated ferromagnetic particles with an absorbed layer of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) diluted by tributyl phosphate (TBP) are being evaluated for application in the separation and the recovery of low concentrations of americium and plutonium from nuclear waste solutions. Due to their chemical nature, these extractants selectively complex americium and plutonium contaminants onto the particles, which can be recovered from the waste solution using a magnet. The effectiveness of the extractant-absorbed particles at removing transuranics (TRU) from simulated solutions and various nitric acid solutions was measured by gamma and liquid scintillation counting of plutonium and americium. The HNO 3 concentration range was 0.01 M to 6M. The partition coefficients (K d ) for various actinides at 2M HNO 3 were determined to be between 3,000 and 30,000. These values are larger than those projected for TRU recovery by traditional liquid/liquid extraction. Results from transmission electron microscopy indicated a large dependence of K d on relative magnetite location within the polymer and the polymer surface area. Energy disperse spectroscopy demonstrated homogeneous metal complexation on the polymer surface with no metal clustering. The radiolytic stability of the particles was determined by using 60 Co gamma irradiation under various conditions. The results showed that K d more strongly depends on the nitric acid dissolution rate of the magnetite than the gamma irradiation dose. Results of actinide separation from simulated high-level waste representative of that at various DOE sites are also discussed

Determination of difficult to measure actinides in radioactive liquid waste

International Nuclear Information System (INIS)

Drabova, V.; Galanda, D.; Dulanska, S.; Remenec, B.; Kuruc, J.

2014-01-01

In decommissioning of a nuclear facilities and radioactive waste treatment the activity of various radionuclides need to be measured for the waste characterization. Radiochemical separation of 241 Am, 237 Np and isotopes of plutonium was tested on model solution of evaporator concentrate sample for isolation of each of them for alpha-spectrometry analysis. This paper describes use of the molecular recognition technology product AnaLig(R)Pu-01 gel from IBC Advanced technologies, Inc. to effectively and selectively pre-concentrate, separate and recover difficult-to-measure actinides from model solution of evaporator concentrate samples which belong to the most difficult matrices to analyse. The method is suitable for analysing highly contaminated samples of radioactive waste in a relatively short time. For counting the alpha activity of 241 Am, 239,240 Pu, 238 Pu and 237 Np ORTEC 576A alpha-spectrometer equipped with ULTRA TM ion implanted silicon detectors (600 mm 2 active area) was used. The spectra were processed by using the Alpha-vision TM 32-bit emulation software from the EG and G ORTEC company. (authors)

Investigation into complexing of pentavalent actinide forms with some anions of organic acids by the coprecipitation method

International Nuclear Information System (INIS)

Moskvin, A.I.; Poznyakov, A.N.; AN SSSR, Moscow. Inst. Geokhimii i Analiticheskoj Khimii)

1979-01-01

Complexing of pentavolent forms of Np, Pu, Am actinides with anions of acetic, oxalic acids and EDTA is studied using the method of coprecipitation with iron hydroxide. Composition and stability constants of the actinide complexes formed are determined. The acids anions are arranged in a row in the order of decrease of complexing tendency that is EDTA anion>C 2 O 4 2- >CH 3 COO -

Institute of Energy and Climate Research IEK-6. Nuclear Waste Management report 2011/2012. Material science for nuclear waste management

International Nuclear Information System (INIS)

Klinkenberg, M.; Neumeier, S.; Bosbach, D.

2013-01-01

The nuclear waste management section of the Institute of Energy and Climate Research IEK-6 in Juelich is focused on research on radiochemistry aspects/materials science relevant for the long-term safety of nuclear waste storage and disposal. Studies on innovative waste management strategies include partitioning o actinides and the development of ceramic waste forms. Structural research is covering solid state chemistry, crystallography and computational science to model actinide containing compounds. With respect to waste management concepts nondestructive essay techniques, waste treatment procedures and product quality control strategies were developed.

Densified waste form and method for forming

Science.gov (United States)

Garino, Terry J.; Nenoff, Tina M.; Sava Gallis, Dorina Florentina

2015-08-25

Materials and methods of making densified waste forms for temperature sensitive waste material, such as nuclear waste, formed with low temperature processing using metallic powder that forms the matrix that encapsulates the temperature sensitive waste material. The densified waste form includes a temperature sensitive waste material in a physically densified matrix, the matrix is a compacted metallic powder. The method for forming the densified waste form includes mixing a metallic powder and a temperature sensitive waste material to form a waste form precursor. The waste form precursor is compacted with sufficient pressure to densify the waste precursor and encapsulate the temperature sensitive waste material in a physically densified matrix.

Safe actinide disposition in molten salt reactors

International Nuclear Information System (INIS)

Gat, U.

1997-01-01

Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs

Crystalline Ceramic Waste Forms: Comparison Of Reference Process For Ceramic Waste Form Fabrication

Energy Technology Data Exchange (ETDEWEB)

Brinkman, K. S. [Savannah River National Laboratory; Marra, J. C. [Savannah River National Laboratory; Amoroso, J. [Savannah River National Laboratory; Tang, M. [Los Alamos National Laboratory

2013-08-22

The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO{sub 2} and Al{sub 2}O{sub 3} were observed distributed in a network of fine grains with small residual pores

Recycling of actinides and nuclear waste products. Annual report from the research programme 1996

International Nuclear Information System (INIS)

Konings, R.J.M.; Bakker, K.; Dodd, D.H.; Gruppelaar, H.; De Haas, J.B.M.; Kloosterman, J.L.

1997-07-01

The research program on the title subject started in 1994 and is planned to be completed in 1998. In this period several technical and scientific aspects of recycling and transmutation are investigated in different projects. The results of the 1996 projects are summarized and described in this report. The 1996 projects concern (1) the chemistry of actinides and inert matrices to test and characterize the matrices and actinide compounds in order to develop uranium-free fissionable materials for the transmutation of actinides; (2) the transmutation of plutonium in light water reactors (LWR) to assess and increase the burnup of plutonium and to assess the safety of plutonium transmutation in LWRs; (3) the radiological consequences of different nuclear fuel cycles; (4) and a reactor physics analysis of new thorium-based reactor systems to study the possibility to reduce the amount of long-living radioactive waste materials by means of the use of thorium-based compounds in a high-temperature reactor (HTR) or accelerators. 15 figs., 6 tabs., 23 refs

Recycling of actinides and nuclear waste products. Annual report of the research programme 1997

International Nuclear Information System (INIS)

Konings, R.J.M.; Bakker, K.; Boerrigter, H.; Damen, P.G.M.; Gruppelaar, H.; Huntelaar, M.E.; Kloosterman, J.L.

1998-07-01

The research program on the title subject started in 1994 and is planned to be completed in 1998. In this period several technical and scientific aspects of recycling and transmutation are investigated in different projects. The results of the 1997 projects, carried out in the period July 1997 to June 1998, are summarized and described in this report. The 1997 projects concern (1) transmutation of actinides in inert matrices with the aim to design, test and characterize uranium-free fission materials for the transmutation of actinides, both for single as for multiple recycling strategies; (2) scenario studies for plutonium recycling with the aim to gain insight into the possibilities to reduce plutonium reserves by using plutonium as a fissionable material in reactors; (3) transmutation by means of accelerator-driven systems with the aim to analyze the options for the burning of plutonium in accelerator-driven, thorium-based systems; and (4) separation of actinides and lanthanides by means of Supported Liquid Membranes with the aim to study the possibility to extract americium from nuclear waste materials. refs

Comparison of different options for minor actinide transmutation in the frame of the French law for waste management

International Nuclear Information System (INIS)

Chabert, Christine; Leudet, Alain; Saturnin, Anne

2011-01-01

In the frame of the French Act for waste management which has been passed by French Parliament on June 28th, 2006, it is requested to obtain in 2012 an assessment of industrial perspectives of partitioning and transmutation of long-lived elements. These studies must be carried out in tight connection with GENIV systems development. The expected results must include the evaluation of technical and economic scenarios taking into account the optimization options between the minor actinide transmutation processes, their interim storage and geological disposal, including an analysis of several criteria. In this perspective, the CEA has established a working group named 'GT TES' (Working Group on Technical and Economic Scenarios) involving EDF and AREVA to define scenarios, the various criteria to evaluate them, to conduct these evaluations and then to highlight the key results. The group also relied on ANDRA for the geological storage studies. The scenarios evaluations take place in the French context. The nuclear energy production is supposed to remain constant during the scenarios and equal to 430 TWhe/year in accordance with the current French nuclear power installed capacity of 60 GW(e). The deployment of the first Sodium-cooled Fast Reactor (SFR) starts in 2040, considering that at this date the SFR technology should be mature. Several management schemes of minor actinides have been studied: Plutonium recycling in SFR (minor actinides are sent to the waste). Plutonium recycling and minor actinide (or Am alone) transmutation in SFR and in homogeneous mode ('Hom.'). Plutonium recycling and minor actinide (or Am alone) transmutation in SFR and in heterogeneous mode ('Het.'). Plutonium recycling in SFR and minor actinide transmutation in Accelerator-Driven-System (ADS). The criteria used to analyze these different scenarios, should take into account the viewpoint of scientists, industrials, administrations, and the general public. They are listed below: Inventories and

Systematics of criticality properties of actinide nuclides and its bearing on the long lived fission waste problem

International Nuclear Information System (INIS)

Srinivasan, M.; Rao, K.S.; Garg, S.B.; Iyengar, P.K.

1989-01-01

This paper reports on a systematic analysis of the criticality parameters of over twenty fissile and fertile isotopes of eight transthorium actinide elements that has been carried out by us. It is observed that K ∞ increases and critical mass decreases monotonically with the fissility parameter (Z 2 /A) of the nuclides. This implies that each and every isotope of transuranic elements such as Np, Am, Cm etc. which are produced as by-products during reactor operation is a more valuable nuclear fuel than the corresponding fissile/fissible isotopes of plutonium. This finding has a profound bearing on the long lived fission waste problem and supports the view that the byproduct actinide elements should be separated from the high level waste stream and recycled back into fission reactors, thereby eliminating one of the commonly voiced concerns regarding the acceptability of nuclear fission power

Glass-ceramic nuclear waste forms obtained by crystallization of SiO 2-Al 2O 3-CaO-ZrO 2-TiO 2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th): Study of the crystallization from the surface

Science.gov (United States)

Loiseau, P.; Caurant, D.

2010-07-01

Glass-ceramic materials containing zirconolite (nominally CaZrTi 2O 7) crystals in their bulk can be envisaged as potential waste forms for minor actinides (Np, Am, Cm) and Pu immobilization. In this study such matrices are synthesized by crystallization of SiO 2-Al 2O 3-CaO-ZrO 2-TiO 2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th) as surrogates. A thin partially crystallized layer containing titanite and anorthite (nominally CaTiSiO 5 and CaAl 2Si 2O 8, respectively) growing from glass surface is also observed. The effect of the nature and concentration of surrogates on the structure, the microstructure and the composition of the crystals formed in the surface layer is presented in this paper. Titanite is the only crystalline phase able to significantly incorporate trivalent lanthanides whereas ThO 2 precipitates in the layer. The crystal growth thermal treatment duration (2-300 h) at high temperature (1050-1200 °C) is shown to strongly affect glass-ceramics microstructure. For the system studied in this paper, it appears that zirconolite is not thermodynamically stable in comparison with titanite growing form glass surface. Nevertheless, for kinetic reasons, such transformation (i.e. zirconolite disappearance to the benefit of titanite) is not expected to occur during interim storage and disposal of the glass-ceramic waste forms because their temperature will never exceed a few hundred degrees.

Crystalline matrices for the immobilization of plutonium and actinides

Energy Technology Data Exchange (ETDEWEB)

Anderson, E.B.; Burakov, E.E.; Galkin, Ya.B.; Starchenko, V.A.; Vasiliev, V.G. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation)

1996-05-01

The management of weapon plutonium, disengaged as a result of conversion, is considered together with the problem of the actinide fraction of long-lived high level radioactive wastes. It is proposed to use polymineral ceramics based on crystalline host-phases: zircon ZrSiO{sub 4} and zirconium dioxide ZrO{sub 2}, for various variants of the management of plutonium and actinides (including the purposes of long-term safe storage or final disposal from the human activity sphere). It is shown that plutonium and actinides are able to form with these phases on ZrSiO{sub 4} and ZrO{sub 2} was done on laboratory level by the hot pressing method, using the plasmochemical calcination technology. To incorporate simulators of plutonium into the structure of ZrSiO{sub 4} and ZrO{sub 2} in the course of synthesis, an original method developed by the authors as a result of studying the high-uranium zircon (Zr,U) SiO{sub 4} form Chernobyl {open_quotes}lavas{close_quotes} was used.

A worldwide perspective on actinide burning

International Nuclear Information System (INIS)

Burch, W.D.

1991-01-01

Worldwide interest has been evident over the past few years in reexamining the merits of recovering the actinides from spent light-water reactor (LWR) fuel and transmuting them in fast reactors to reduce hazards in geologic repositories. This paper will summarize some of the recent activities in this field. Several countries are embarked on programs of reprocessing and vitrification of present wastes, from which removal of the actinides is largely precluded. The United States is assessing the ideas related to the fast reactor program and the potential application to defense wastes. 18 refs., 2 figs

Preliminary minimum detectable limit measurements in 208-L drums for selected actinide isotopes in mock-waste matrices

International Nuclear Information System (INIS)

Camp, D.C.; Wang, Tzu-Fang; Martz, H.E.

1992-01-01

Preliminary minimum detectable levels (MDLS) of selected actinide isotopes have been determined in full-scale, 55-gallon drums filled with a range of mock-waste materials from combustibles (0.14 g/CM 3 ) to sand (1.7 g/CM 3 ). Measurements were recorded from 100 to 10,000 seconds with selected actinide sources located in these drums at an edge position, on the center axis of a drum and midway between these two positions. Measurements were also made with a 166 Ho source to evaluate the attenuation of these mock-matrix materials as a function of energy. By knowing where the source activity is located within a drum, our preliminary results show that a simply collimated 90% HPGE detector can differentiate between TRU (>100 nCi/g) and LLW amounts of 239 Pu in only 100s of measurement time and with sufficient accuracy in both low and medium density, low Z materials. Other actinides measured so far include 235 U, 241 Am, and 244 Cm. These measurements begin to establish the probable MDLs achievable in the nondestructive assays of real waste drums when using active and passive CT. How future measurements may differ from these preliminary measurements is also discussed

Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

International Nuclear Information System (INIS)

Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

2009-01-01

Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

Impact of minor actinide recycling on sustainable fuel cycle options

Energy Technology Data Exchange (ETDEWEB)

Heidet, F.; Kim, T. K.; Taiwo, T. A.

2017-11-01

The recent Evaluation and Screening study chartered by the U.S. Department of Energy, Office of Nuclear Energy, has identified four fuel cycle options as being the most promising. Among these four options, the two single-stage fuel cycles rely on a fast reactor and are differing in the fact that in one case only uranium and plutonium are recycled while in the other case minor actinides are also recycled. The two other fuel cycles are two-stage and rely on both fast and thermal reactors. They also differ in the fact that in one case only uranium and plutonium are recycled while in the other case minor actinides are also recycled. The current study assesses the impact of recycling minor actinides on the reactor core design, its performance characteristics, and the characteristics of the recycled material and waste material. The recycling of minor actinides is found not to affect the reactor core performance, as long as the same cycle length, core layout and specific power are being used. One notable difference is that the required transuranics (TRU) content is slightly increased when minor actinides are recycled. The mass flows are mostly unchanged given a same specific power and cycle length. Although the material mass flows and reactor performance characteristics are hardly affected by recycling minor actinides, some differences are observed in the waste characteristics between the two fuel cycles considered. The absence of minor actinides in the waste results in a different buildup of decay products, and in somewhat different behaviors depending on the characteristic and time frame considered. Recycling of minor actinides is found to result in a reduction of the waste characteristics ranging from 10% to 90%. These results are consistent with previous studies in this domain and depending on the time frame considered, packaging conditions, repository site, repository strategy, the differences observed in the waste characteristics could be beneficial and help improve

Actinide burning in the integral fast reactor

International Nuclear Information System (INIS)

Chang, Y.I.

1993-01-01

During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the IFR concept is its unique pyroprocessing. Pyroprocessing has the potential to radically improve long-term waste management strategies by exploiting the following attributes: 1. Minor actinides accompany plutonium product stream; therefore, actinide recycling occurs naturally. Actinides, the primary source of long-term radiological toxicity, are removed from the waste stream and returned to the reactor for in situ burning, generating useful energy. 2. High-level waste volume from pyroprocessing call be reduced substantially as compared with direct disposal of spent fuel. 3. Decay heat loading in the repository can be reduced by a large factor, especially for the long-term burden. 4. Low-level waste generation is minimal. 5. Troublesome fission products, such as 99 Tc, 129 I, and 14 C, are contained and immobilized. Singly or in combination, the foregoing attributes provide important improvements in long-term waste management in terms of the ease in meeting technical performance requirements (perhaps even the feasibility of demonstrating that technical performance requirements can be met) and perhaps also in ultimate public acceptance. Actinide recycling, if successfully developed, could well help the current repository program by providing an opportunity to enhance capacity utilization and by deferring the need for future repositories. It also represents a viable technical backup option in the event unforeseen difficulties arise in the repository licensing process

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

International Nuclear Information System (INIS)

1986-01-01

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

NEARSOL, Aqueous Speciation and Solubility of Actinides for Waste Disposal

International Nuclear Information System (INIS)

Leach, S.J.; Pryke, D.C.

1989-01-01

A - Description of program or function: NEARSOL models the aqueous speciation and solubility of actinides under near-field conditions for disposal using a simple thermodynamic approach. B - Method of solution: The program draws information from a thermodynamic data base consisting of solubility products and complex formation constants for all known species, and standard electrode potentials, at 25 C, corrected for ionic strength effects. By minimising the free energy of the system through a series of iterations, a precipitating solid phase is predicted which limits the solubility, and the concentration of the main aqueous species are calculated as a function of pH. Initially the program evaluates only hydroxide and carbonate species, but the effect of sulphate, phosphate and fluoride anions can also be included. The program is simple to use, requiring inputs of: 1. Actinide(s); 2. pH range; 3. Ionic strength; 4. Redox conditions; 5. Ligand concentrations. Functions are included to calculate the distribution of the protonated and un-protonated forms of carbonate and phosphate and the value of Eh as a function of pH under disposal conditions as required. The program can further evaluate the role of free calcium ions. C - Restrictions on the complexity of the problem: None

Recovery of actinides from actinide-aluminium alloys by chlorination: Part I

Energy Technology Data Exchange (ETDEWEB)

Cassayre, L., E-mail: cassayre@chimie.ups-tlse.fr [Laboratoire de Genie Chimique (LGC), Departement Procedes Electrochimiques, CNRS-UMR 5503, Universite de Toulouse III - Paul Sabatier, 31062 Toulouse (France); Soucek, P.; Mendes, E.; Malmbeck, R.; Nourry, C.; Eloirdi, R.; Glatz, J.-P. [European Commission, JRC, Institute for Transuranium Elements, Postfach 2340, 76125 Karlsruhe (Germany)

2011-07-01

Pyrochemical processes in molten LiCl-KCl are being developed in ITU for recovery of actinides from spent nuclear fuel. The fuel is anodically dissolved to the molten salt electrolyte and actinides are electrochemically reduced on solid aluminium cathodes forming solid actinide-aluminium alloys. A chlorination route is being investigated for recovery of actinides from the alloys. This route consists in three steps: Vacuum distillation for removal of the salt adhered on the electrode, chlorination of the actinide-aluminium alloys by chlorine gas and sublimation of the formed AlCl{sub 3}. A thermochemical study showed thermodynamic feasibility of all three steps. On the basis of the conditions identified by the calculations, experiments using pure UAl{sub 3} alloy were carried out to evaluate and optimise the chlorination step. The work was focused on determination of the optimal temperature and Cl{sub 2}/UAl{sub 3} molar ratio, providing complete chlorination of the alloy without formation of volatile UCl{sub 5} and UCl{sub 6}. The results showed high efficient chlorination at a temperature of 150 deg. C.

Recovery of actinides from actinide-aluminium alloys by chlorination: Part I

International Nuclear Information System (INIS)

Cassayre, L.; Soucek, P.; Mendes, E.; Malmbeck, R.; Nourry, C.; Eloirdi, R.; Glatz, J.-P.

2011-01-01

Pyrochemical processes in molten LiCl-KCl are being developed in ITU for recovery of actinides from spent nuclear fuel. The fuel is anodically dissolved to the molten salt electrolyte and actinides are electrochemically reduced on solid aluminium cathodes forming solid actinide-aluminium alloys. A chlorination route is being investigated for recovery of actinides from the alloys. This route consists in three steps: Vacuum distillation for removal of the salt adhered on the electrode, chlorination of the actinide-aluminium alloys by chlorine gas and sublimation of the formed AlCl 3 . A thermochemical study showed thermodynamic feasibility of all three steps. On the basis of the conditions identified by the calculations, experiments using pure UAl 3 alloy were carried out to evaluate and optimise the chlorination step. The work was focused on determination of the optimal temperature and Cl 2 /UAl 3 molar ratio, providing complete chlorination of the alloy without formation of volatile UCl 5 and UCl 6 . The results showed high efficient chlorination at a temperature of 150 deg. C.

Environmental chemistry of the actinide elements

International Nuclear Information System (INIS)

Rao Linfeng

1986-01-01

The environmental chemistry of the actinide elements is a new branch of science developing with the application of nuclear energy on a larger and larger scale. Various aspects of the environmental chemistry of the actinide elements are briefly reviewed in this paper, such as its significance in the nuclear waste disposal, its coverage of research fields and possible directions for future study

Selective partitioning of mercury from co-extracted actinides in a simulated acidic ICPP waste stream

International Nuclear Information System (INIS)

Brewer, K.N.; Herbst, R.S.; Tranter, T.J.

1995-01-01

The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/l. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, 203 Hg, was spiked into a simulated solution of SBW containing 1 g/l mercury. Successive extraction batch contacts with the mercury spiked waste simulant and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl 2 from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HNO 3 and 0.077 with 0.25 M Na 2 CO 3 . An experimental flowsheet was designed from the batch contact tests and tested counter-currently using 5.5 cm centrifugal contactors. Results from the counter-current test show that mercury can be removed from the acidic mixed SBW simulant and recovered separately from the actinides

A literature review of actinide-carbonate mineral interactions

International Nuclear Information System (INIS)

Stout, D.L.

1993-10-01

Chemical retardation of actinides in groundwater systems is a potentially important mechanism for assessing the performance of the Waste Isolation Pilot Plant (WIPP), a facility intended to demonstrate safe disposal of transuranic waste. Rigorous estimation of chemical retardation during transport through the Culebra Dolomite, a water-bearing unit overlying the WIPP, requires a mechanistic understanding of chemical reactions between dissolved elements and mineral surfaces. This report represents a first step toward this goal by examining the literature for pertinent experimental studies of actinide-carbonate interactions. A summary of existing models is given, along with the types of experiments on which these models are based. Articles pertaining to research into actinide interactions with carbonate minerals are summarized. Select articles involving trace element-carbonate mineral interactions are also reviewed and may serve as templates for future research. A bibliography of related articles is included. Americium(III), and its nonradioactive analog neodymium(III), partition strongly from aqueous solutions into carbonate minerals. Recent thermodynamic, kinetic, and surface studies show that Nd is preferentially removed from solution, forming a Nd-Ca carbonate solid solution. Neptunium(V) is rapidly removed from solution by carbonates. Plutonium incorporation into carbonates is complicated by multiple oxidation states. Little research has been done on the radium(H) and thorium(IV) carbonate systems. Removal of uranyl ion from solution by calcite is limited to monolayer surface coverage

Radioactive decay pattern of actinides present in waste from Mo-99 production

Energy Technology Data Exchange (ETDEWEB)

Hiromoto, Goro; Dellamano, José Claudio, E-mail: hiromoto@ipen.br, E-mail: jcdellam@ipen.br [Instituto de PesquisasEnergéticas e Nucleares (GRR/IPEN/CNEN-SP), São Paulo, SP (Brazil). Gerência de Rejeitos Radioativos

2017-07-01

Brazil is currently planning to produce {sup 99}Mo from fission of LEU targets to meet the present national demand of {sup 99m}Tc. The {sup 99}Mo activity planned at the end of irradiation is 5000 Ci (185 TBq) per weekly cycle, in order to meet the present demand of 1000 Ci (37 TBq) per week, after target cooling and processing. To predict the activities that will be handled in the waste treatment facility, the computational code SCALE 6.0 was used to simulate the irradiation of the uranium targets and the decay of radioactive products. This study presents the findings of this research, mainly focused on the actinides activity that will be present in the waste and the respective radioactive decay pattern over a period of one hundred thousand years. (author)

Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model

Energy Technology Data Exchange (ETDEWEB)

Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

2013-02-01

This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

Actinide partitioning-transmutation program final report. III. Transmutation studies

International Nuclear Information System (INIS)

Wachter, J.W.; Croff, A.G.

1980-07-01

Transmutation of the long-lived nuclides contained in fuel cycle wastes has been suggested as a means of reducing the long-term toxicity of the wastes. A comprehensive program to evaluate the feasibility and incentives for recovering the actinides from wastes (partitioning) and transmuting them to short-lived or stable nuclides has been in progress for 3 years under the direction of Oak Ridge National Laboratory (ORNL). This report constitutes the final assessment of transmutation in support of this program. Included are (1) a summary of recent transmutation literature, (2) a generic evaluation of actinide transmutation in thermal, fast, and other transmutation devices, (3) a preliminary evaluation of 99 Tc and 129 I transmutation, and (4) a characterization of a pressurized-water-reactor fuel cycle with and without provisions for actinide recovery and transmutation for use in other parts of the ORNL program. The principal conclusion of the report is that actinide transmutation is feasible in both thermal and fast reactors, subject to demonstrating satisfactory fuel performance, with relatively little impact on the reactor. It would also appear that additional transmutation studies are unwarranted until a firm decision to proceed with actinide transmutation has been made by the responsible authorities

Incineration of actinide targets in a pressurized water reactor spin project

International Nuclear Information System (INIS)

Puill, A.; Bergeron, J.

1993-01-01

The ability of Pressurized Water Reactors (PWR) with uranium fuel to limit the inventory growth of minor actinides (237 neptunium, and americium) produced by the French nuclear powerplants is studied. Targets containing an actinide oxide mixed to an inert matrix are loaded in some reactors. After being irradiated along with the fuel, the target is specially reprocessed. The remaining actinide and the plutonium which is produced, added to fresh actinide, are recycled in new targets. The radiotoxicity balance, with and without incineration, is examined considering that only the losses coming from the target reprocessing treated as waste. A scenario arbitrarily based on 18 years of operation results in a reduction of the radiotoxicity of the waste by a factor between 10 and 20, depending on the actinide considered. 6 refs., 6 figs., 6 tabs

Design, synthesis, and evaluation of polyhydroxamate chelators for selective complexation of actinides

International Nuclear Information System (INIS)

Gopalan, A.; Jacobs, H.; Koshti, N.; Stark, P.; Huber, V.; Dasaradhi, L.; Caswell, W.; Smith, P.; Jarvinen, G.

1995-01-01

Specific chelating polymers targeted for actinides have much relevance to problems involving remediation of nuclear waste. Goal is to develop polymer supported, ion specific extraction systems for removing actinides and other hazardous metal ions from wastewaters. This is part of an effort to develop chelators for removing actinide ions such as Pu from soils and waste streams. Selected ligands are being attached to polymeric backbones to create novel chelating polymers. These polymers and other water soluble and insoluble polymers have been synthesized and are being evaluated for ability to selectively remove target metal ions from process waste streams

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

Modeling corrosion and constituent release from a metal waste form

International Nuclear Information System (INIS)

Bauer, T. H.; Fink, J. K.; Abraham, D. P.; Johnson, I.; Johnson, S. G.; Wigeland, R. A.

2000-01-01

Several ANL ongoing experimental programs have measured metal waste form (MWF) corrosion and constituent release. Analysis of this data has initiated development of a consistent and quantitative phenomenology of uniform aqueous MWF corrosion. The effort so far has produced a preliminary fission product and actinide release model based on measured corrosion rates and calibrated by immersion test data for a 90 C J-13 and concentrated J-13 solution environment over 1-2 year exposure times. Ongoing immersion tests of irradiated and unirradiated MWF samples using more aggressive test conditions and improved tracking of actinides will serve to further validate, modify, and expand the application base of the preliminary model-including effects of other corrosion mechanisms. Sample examination using both mechanical and spectrographic techniques will better define both the nature and durability of the protective barrier layer. It is particularly important to assess whether the observations made with J-13 solution at 900 C persist under more aggressive conditions. For example, all the multiplicative factors in Table 1 implicitly assume the presence of protective barriers. Under sufficiently aggressive test conditions, such protective barriers may very well be altered or even eliminated

Double liquid membrane system for the removal of actinides and lanthanides from acidic nuclear wastes

International Nuclear Information System (INIS)

Chiarizia, R.; Danesi, P.R.

1985-01-01

Supported liquid membranes (SLM), consisting of an organic solution of n-octyl-(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and tributyl-phosphate (TBP) in decalin are able to perform selective separation and concentration of actinide and lanthanide ions from aqueous nitrate feed solutions and synthetic nuclear wastes. In the membrane process a possible strip solution is a mixture of formic acid and hydroxylammonium formate (HAF). The effectiveness of this strip solution is reduced and eventually nullified by the simultaneous transfer through the SLM of nitric acid which accumulates in the strip solution. A possible way to overcome this drawback is to make use of a second SLM consisting of a primary amine which is able to extract only HNO 3 from the strip solution. In this work the results obtained by experimentally studying the membrane system: synthetic nuclear waste/CMPO-TBP membrane/HCOOH-HAF strip solution/primary amine membrane/NaOH solution, are reported. They show that the use of a second liquid membrane is effective in controlling the HNO 3 concentration in the strip solution, thus allowing the actinide and lanthanide ions removal from the feed solution to proceed to completion. 15 refs., 10 figs., 1 tab

Simulations of the Thermodynamic and Diffusion Properties of Actinide Oxide Fuel Materials

International Nuclear Information System (INIS)

Becker, Udo

2013-01-01

Spent nuclear fuel from commercial reactors is comprised of 95-99 percent UO 2 and 1-5 percent fission products and transuranic elements. Certain actinides and fission products are of particular interest in terms of fuel stability, which affects reprocessing and waste materials. The transuranics found in spent nuclear fuels are Np, Pu, Am, and Cm, some of which have long half- lives (e.g., 2.1 million years for 237 Np). These actinides can be separated and recycled into new fuel matrices, thereby reducing the nuclear waste inventory. Oxides of these actinides are isostructural with UO 2 , and are expected to form solid solutions. This project will use computational techniques to conduct a comprehensive study on thermodynamic properties of actinide-oxide solid solutions. The goals of this project are to: Determine the temperature-dependent mixing properties of actinide-oxide fuels; Validate computational methods by comparing results with experimental results; Expand research scope to complex (ternary and quaternary) mixed actinide oxide fuels. After deriving phase diagrams and the stability of solid solutions as a function of temperature and pressure, the project team will determine whether potential phase separations or ordered phases can actually occur by studying diffusion of cations and the kinetics of potential phase separations or ordered phases. In addition, the team will investigate the diffusion of fission product gases that can also have a significant influence on fuel stability. Once the system has been established for binary solid solutions of Th, U, Np, and Pu oxides, the methodology can be quickly applied to new compositions that apply to ternaries and quaternaries, higher actinides (Am, Cm), burnable poisons (B, Gd, Hf), and fission products (Cs, Sr, Tc) to improve reactivity

IAEA Activities on Assessment of Partitioning Processes for Transmutation of Actinides

International Nuclear Information System (INIS)

Basak, Uddharan; Dyck, Gary R.

2010-01-01

In these days of nuclear renaissance, appropriate management of radioactive materials arising from the nuclear fuel cycle back end is one of the most important issues related to the long term sustainability of nuclear energy. The present practice in the back end of the closed fuel cycle involves the recovery of uranium and plutonium from spent fuel by the aqueous based PUREX process for reuse in reactors and the conditioning of reprocessing waste into a form suitable for long term storage. The waste contains mainly fission products and transuranium elements immobilized in glass matrix. However, advanced fuel cycles incorporating partitioning of actinides along with minor actinides and their subsequent transmutation (P and T) in a fast neutron energy spectrum could be proliferation resistant and at the same time reduce the waste radiotoxicity by many orders of magnitude. Considering the importance of P and T on long term sustainability, the International Atomic Energy Agency has initiated many collaborative research programs in this area as part of our advanced fuel cycle activities. This paper presents the current and future activities on advanced partitioning methods, highlighting the challenges associated with these processes, fuel manufacturing techniques suitable for integration with reprocessing facility and the IAEA's minor actinide data base (MADB), as a part of integrated nuclear fuel cycle information system (iNFCIS). (authors)

Actinide oxide photodiode and nuclear battery

Energy Technology Data Exchange (ETDEWEB)

Sykora, Milan; Usov, Igor

2017-12-05

Photodiodes and nuclear batteries may utilize actinide oxides, such a uranium oxide. An actinide oxide photodiode may include a first actinide oxide layer and a second actinide oxide layer deposited on the first actinide oxide layer. The first actinide oxide layer may be n-doped or p-doped. The second actinide oxide layer may be p-doped when the first actinide oxide layer is n-doped, and the second actinide oxide layer may be n-doped when the first actinide oxide layer is p-doped. The first actinide oxide layer and the second actinide oxide layer may form a p/n junction therebetween. Photodiodes including actinide oxides are better light absorbers, can be used in thinner films, and are more thermally stable than silicon, germanium, and gallium arsenide.

Research on the actinide chemistry in Nuclear Fuel Cycle

Energy Technology Data Exchange (ETDEWEB)

Song, Kyseok; Park, Yong Joon; Cho, Young Hwan; and others

2012-04-15

Fundamental technique to measure chemical behaviors and properties of lanthanide and actinide in radioactive waste is necessary for the development of pryochemical process. First stage, the electrochemical/spectroscopic integrated measurement system was designed and set up for spectro-electrochemical measurements of lanthanide and actinide ions in high temperature molten salt media. A compact electrochemical cell and electrode system was also developed for the minimization of reactants, and consequently minimization of radioactive waste generation. By applying these equipments, oxidation and reduction behavior of lanthanide and actinide ions in molten salt media have been made. Also, thermodynamic parameter values are determined by interpreting the results obtained from electrochemical measurements. Several lanthanide ions exhibited fluorescence properties in molten salt. Also, UV-VIS measurement provided the detailed information regarding the oxidation states of lanthanide and actinide ions in high temperature molten salt media. In the second stage, measurement system for physical properties at pyrochemical process such as viscosity, melting point and conductivity is established, and property database at different compositions of lanthanide and actinide is collected. And, both interactions between elements and properties with different potential are measured at binary composition of actinide-lanthanide in molten salt using electrochemical/spectroscopic integrated measurement system.

Investigation of microscopic radiation damage in waste forms using ODNMR and AEM techniques. (EMSP Project Final Report)

International Nuclear Information System (INIS)

Liu, G.; Luo, J.; Beitz, J.; Li, S.; Williams, C.; Zhorin, V.

2000-01-01

This project seeks to understand the microscopic effects of radiation damage in nuclear waste forms. The authors' approach to this challenge encompasses studies of ceramics and glasses containing short-lived alpha- and beta-emitting actinides with electron microscopy, laser and X-ray spectroscopic techniques, and computational modeling and simulations. In order to obtain information on long-term radiation effects on waste forms, much of the effort is to investigate α-decay induced microscopic damage in 18-year old samples of crystalline yttrium and lutetium orthophosphates that initially contained ∼ 1(wt)% of the alpha-emitting isotope 244 Cm (18.1 y half life). Studies also are conducted on borosilicate glasses that contain 244 Cm, 241 Am, or 249 Bk, respectively. The authors attempt to gain clear insights into the properties of radiation-induced structure defects and the consequences of collective defect-environment interactions, which are critical factors in assessing the long-term performance of high-level nuclear waste forms

Investigation of microscopic radiation damage in waste forms using ODNMR and AEM techniques. (EMSP Project Final Report)

Energy Technology Data Exchange (ETDEWEB)

Liu, G.; Luo, J.; Beitz, J.; Li, S.; Williams, C.; Zhorin, V.

2000-04-21

This project seeks to understand the microscopic effects of radiation damage in nuclear waste forms. The authors' approach to this challenge encompasses studies of ceramics and glasses containing short-lived alpha- and beta-emitting actinides with electron microscopy, laser and X-ray spectroscopic techniques, and computational modeling and simulations. In order to obtain information on long-term radiation effects on waste forms, much of the effort is to investigate {alpha}-decay induced microscopic damage in 18-year old samples of crystalline yttrium and lutetium orthophosphates that initially contained {approximately} 1(wt)% of the alpha-emitting isotope {sup 244}Cm (18.1 y half life). Studies also are conducted on borosilicate glasses that contain {sup 244}Cm, {sup 241}Am, or {sup 249}Bk, respectively. The authors attempt to gain clear insights into the properties of radiation-induced structure defects and the consequences of collective defect-environment interactions, which are critical factors in assessing the long-term performance of high-level nuclear waste forms.

Extraction processes and solvents for recovery of cesium, strontium, rare earth elements, technetium and actinides from liquid radioactive waste

Science.gov (United States)

Zaitsev, Boris N.; Esimantovskiy, Vyacheslav M.; Lazarev, Leonard N.; Dzekun, Evgeniy G.; Romanovskiy, Valeriy N.; Todd, Terry A.; Brewer, Ken N.; Herbst, Ronald S.; Law, Jack D.

2001-01-01

Cesium and strontium are extracted from aqueous acidic radioactive waste containing rare earth elements, technetium and actinides, by contacting the waste with a composition of a complex organoboron compound and polyethylene glycol in an organofluorine diluent mixture. In a preferred embodiment the complex organoboron compound is chlorinated cobalt dicarbollide, the polyethylene glycol has the formula RC.sub.6 H.sub.4 (OCH.sub.2 CH.sub.2).sub.n OH, and the organofluorine diluent is a mixture of bis-tetrafluoropropyl ether of diethylene glycol with at least one of bis-tetrafluoropropyl ether of ethylene glycol and bis-tetrafluoropropyl formal. The rare earths, technetium and the actinides (especially uranium, plutonium and americium), are extracted from the aqueous phase using a phosphine oxide in a hydrocarbon diluent, and reextracted from the resulting organic phase into an aqueous phase by using a suitable strip reagent.

Methods of high-sensitive analysis of actinides in liquid radioactive waste

International Nuclear Information System (INIS)

Diakov, Alexandre A.; Perekhozheva, Tatiana N.; Zlokazova, Elena I.

2002-01-01

A complex of methods has been developed to determine actinides in liquid radioactive wastes for solving the problems of radiation, nuclear and ecological safety of nuclear reactors. The main method is based on the radiochemical separation of U, Np-Pu, Am-Cm on ion-exchange and extraction columns. An identification of radionuclides and determination of their content are performed using alpha-spectrometry. The microconcentrations of the sum of the main fissile materials U-235 and Pu-239 are determined with the usage of plastic track detectors. An independent method of U-238 content determination is the neutron activation analysis. Am-241 content is possible to determine with gamma-spectrometry. (author)

Leaching of actinides from simulated nuclear waste glass

International Nuclear Information System (INIS)

Pickering, S.; Walker, C.T.; Offermann, P.

1982-01-01

Two types of simulated nuclear waste glass doped with actinides were leached at 200 0 C in distilled water and salt solutions. Am, Np, Pu and U were all preferentially retained in the surface layer on the glass. Leaching ratios of 0.1 to 0.2 for Np and approx. 0.02 for Am were measured. The losses of Am and Np to the leachant were proportional to the total weight loss of the glass and were larger at 10 ml leachant/cm 2 glass than at 5 ml/cm 2 . Weight loss from the glass occurred only at the start of the experiments for periods ranging from 10 h to 10 days according to leachant composition and volume. Wt losses from the C31-3-EC glass were much greater in saturated NaCl solution than in distilled water. Enrichment in the outer surface layer of Al or Ca according to glass type could be correlated with leachant pH, glass composition and weight loss measurements

Casting of metallic fuel containing minor actinide additions

International Nuclear Information System (INIS)

Trybus, C.L.; Henslee, S.P.; Sanecki, J.E.

1992-01-01

A significant attribute of the Integral Fast Reactor (IFR) concept is the transmutation of long-lived minor actinide fission products. These isotopes require isolation for thousands of years, and if they could be removed from the waste, disposal problems would be reduced. The IFR utilizes pyroprocessing of metallic fuel to separate auranium, plutonium, and the minor actinides from nonfissionable constituents. These materials are reintroduced into the fuel and reirradiated. Spent IFR fuel is expected to contain low levels of americium, neptunium, and curium because the hard neutron spectrum should transmute these isotopes as they are produced. This opens the possibility of using an IFR to trnasmute minor actinide waste from conventional light water reactors (LWRs). A standard IFR fuel is based on the alloy U-20% Pu-10% Zr (in weight percent). A metallic fuel system eases the requirements for reprocessing methods and enables the minor actinide metals to be incorporated into the fuel with simple modifications to the basic fuel casting process. In this paper, the authors report the initial casting experience with minor actinide element addition to an IFR U-Pu-Zr metallic fuel

Interim waste storage for the Integral Fast Reactor

International Nuclear Information System (INIS)

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

Actinide extraction from ICPP sodium bearing waste with 0.75 M DHDECMP/TBP in Isopar L reg-sign

International Nuclear Information System (INIS)

Herbst, R.S.; Brewer, K.N.; Garn, T.G.; Law, J.D.; Rodriguez, A.M.; Tillotson, R.T.

1996-01-01

Recent process development efforts at the Idaho Chemical Processing Plant include examination of solvent extraction technologies for actinide partitioning from sodium bearing waste (SBW) solutions. The use of 0.75 M dihexyl-N, N-diethylcarbamoylmethylphosphonate (DHDECMP or simply CMP) and 1.0 M tri-n-butyl phosphate (TBP) diluted in Isopar L reg-sign was explored for actinide removal from simulated SBW solutions. Experimental evaluations included batch contacts in radiotracer tests with simulated sodium bearing waste solution to measure the extraction and recovery efficiency of the organic solvent. The radioactive isotopes utilized for this study included Pu-238, Pu-239, Am-241, U-233, Np-239, Zr-95, Tc-99m, and Hg-203. Extraction contacts of the organic solvent with the traced SBW stimulant, strip (back-extraction) contacts of the loaded organic solvent with either a 1-hydroxyethane-1, 1-diphosphonic acid (HEDPA) in nitric acid solution or an oxalic acid in nitric acid solution, and solvent wash contacts with sodium carbonate were performed

Mathematical modeling of the effects of aerobic and anaerobic chelate biodegradation on actinide speciation

International Nuclear Information System (INIS)

Banaszak, J.E.; VanBriesen, J.; Rittmann, B.E.; Reed, D.T.

1998-01-01

Biodegradation of natural and anthropogenic chelating agents directly and indirectly affects the speciation, and, hence, the mobility of actinides in subsurface environments. We combined mathematical modeling with laboratory experimentation to investigate the effects of aerobic and anaerobic chelate biodegradation on actinide [Np(IV/V), Pu(IV)] speciation. Under aerobic conditions, nitrilotriacetic acid (NTA) biodegradation rates were strongly influenced by the actinide concentration. Actinide-chelate complexation reduced the relative abundance of available growth substrate in solution and actinide species present or released during chelate degradation were toxic to the organisms. Aerobic bio-utilization of the chelates as electron-donor substrates directly affected actinide speciation by releasing the radionuclides from complexed form into solution, where their fate was controlled by inorganic ligands in the system. Actinide speciation was also indirectly affected by pH changes caused by organic biodegradation. The two concurrent processes of organic biodegradation and actinide aqueous chemistry were accurately linked and described using CCBATCH, a computer model developed at Northwestern University to investigate the dynamics of coupled biological and chemical reactions in mixed waste subsurface environments. CCBATCH was then used to simulate the fate of Np during anaerobic citrate biodegradation. The modeling studies suggested that, under some conditions, chelate degradation can increase Np(IV) solubility due to carbonate complexation in closed aqueous systems

Mathematical modelling of the effects of aerobic and anaerobic chelate biodegradation on actinide speciation

International Nuclear Information System (INIS)

Banaszak, J.E.; VanBriesen, J.M.; Rittmann, B.E.; Reed, D.T.

1998-01-01

Biodegradation of natural and anthropogenic chelating agents directly and indirectly affects the speciation, and hence, the mobility of actinides in subsurface environments. We combined mathematical modelling with laboratory experimentation to investigate the effects of aerobic and anaerobic chelate biodegradation on actinide [Np(IV/V), Pu(IV)] speciation. Under aerobic conditions, nitrilotriacetic acid (NTA) biodegradation rates were strongly influenced by the actinide concentration. Actinide-chelate complexation reduced the relative abundance of available growth substrate in solution and actinide species present or released during chelate degradation were toxic to the organisms. Aerobic bioutilization of the chelates as electron-donor substrates directly affected actinide speciation by releasing the radionuclides from complexed form into solution, where their fate was controlled by inorganic ligands in the system. Actinide speciation was also indirectly affected by pH changes caused by organic biodegradation. The two concurrent processes of organic biodegradation and actinide aqueous chemistry were accurately linked and described using CCBATCH, a computer model developed at Northwestern University to investigate the dynamics of coupled biological and chemical reactions in mixed waste subsurface environments. CCBATCH was then used to simulate the fate of Np during anaerobic citrate biodegradation. The modelling studies suggested that, under some conditions, chelate degradation can increase Np(IV) solubility due to carbonate complexation in closed aqueous systems. (orig.)

Synthesis and characterization of novel lanthanide- and actinide-containing titanates and zircono-titanates; relevance to nuclear waste disposal

International Nuclear Information System (INIS)

Shoup, S.L.S.

1995-08-01

Before experiments using actinide elements are performed, synthetic routes are tested using lanthanides of comparable ionic radii as surrogates. Compound and solid solution formation in several lanthanide-containing titanate and zircono-titanate systems have been established using X-ray diffraction (XRD) analysis, which helped to define interesting and novel experiments, some of which have been performed and are discussed, for selected actinide elements. The aqueous solubilities of several lanthanide- and actinide-containing compounds, representative of the systems studied, were tested in several leachants, including the WIPP open-quotes Aclose quotes brine, following modified Materials Characterization Center procedures (MCC-3). The WIPP open-quotes Aclose quotes brine is a synthetic substitute for that found in nature at the Waste Isolation Pilot Plant (WIPP) in New Mexico. The concentrations of cerium, used as a surrogate for plutonium, leached by the WIPP open-quotes Aclose quotes brine from all the cerium-containing compounds and solid solutions tested were below the Inductively Coupled Plasma (ICP) atomic emission spectrometry limit of detection (10 ppm) established for cerium in this brine. The concentrations of plutonium leached from the two plutonium-containing solid solutions were less than 1 ppm as determined by gross alpha counting and alpha pulse height analysis. Concentrations of strontium leached by the WIPP brine from stable strontium containing titanate compounds, studied as possible immobilizers of both 90 Sr and actinide elements, were also quite low. These compound and solid solution formation investigations and the aqueous solubility studies suggest that the types of titanate and zircono-titanate compounds and solid solutions studied in this work appear to be useful as host matrices for nuclear waste immobilization

Treatment of liquid waste containing alpha nuclides by adsorption

International Nuclear Information System (INIS)

Zeng Jishu; Su Xiguang; Xia Dejing; Fan Sianhua

1997-01-01

In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10 3 Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs

Leaching of actinides and technetium from simulated high-level waste glass

International Nuclear Information System (INIS)

Bradley, D.J.; Harvey, C.O.; Turcotte, R.P.

1979-08-01

Leach tests were conducted using a modified version of the IAEA procedure to study the behavior of glass waste-solution interactions. Release rates were determined for Tc, U, Np, Pu, Am, Cm, and Si in the following solutions: WIPP B salt brine, NaCl (287 g/l), NaCl (1.76 g/1), CaCl 2 (1.66 g/l), NaHCO 3 (2.52 g/l), and deionized water. The leach rates for all elements decreased an order of magnitude from their initial values during the first 20 to 30 days leaching time. The sodium bicarbonate solution produced the highest elemental release rates, while the saturated salt brine and deionized water in general gave the lowest release. Technetium has the highest initial release of all elements studied. The technetium release rates, however, decreased by over four orders of magnitude in 150 days of leaching time. In the prepared glass, technetium was phase separated, concentrating on internal pore surfaces. Neptunium, in all cases except CaCl 2 solution, shows the highest actinide release rate. In general, curium and uranium have the lowest release rates. The range of actinide release rates is from 10 -5 to 10 -8 g/cm 2 /day. 25 figures, 7 tables

Actinides burnup in a sodium fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2017-09-15

The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

International Nuclear Information System (INIS)

Ackerman, J.P.; Johnson, T.R.; Laidler, J.J.

1994-01-01

Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix

Actinide recycling in reactors

International Nuclear Information System (INIS)

Kuesters, H.; Wiese, H.W.; Krieg, B.

1995-01-01

The objective is an assessment of the transmutation of long-lived actinides and fission products and the incineration of plutonium for reducing the risk potential of radioactive waste from reactors in comparison to direct waste disposal. The contribution gives an interim account on homogeneous and heterogeneous recycling of 'risk nuclides' in thermal and fast reactors. Important results: - A homogeneous 5 percent admixture of minor actinides (MA) from N4-PWRs to EFR fuel would allow a transmutation not only of the EFR MA, but in addition of the MA from 5 or 6 PWRs of equal power. However, the incineration is restricted by safety considerations. - LWR have only a very low MA incineration potential, due to their disadvantageous neutron capture/fission ratio. - In order to keep the Cm inventory at a low level, it is advantageous to concentrate the Am heterogeneously in particular fuel elements or rods. (orig./HP)

Actinide science. Fundamental and environmental aspects

International Nuclear Information System (INIS)

Choppin, Gregory R.

2005-01-01

Nuclear test explosions and reactor wastes have deposited an estimated 16x10 15 Bq of plutonium into the world's aquatic systems. However, plutonium concentration in open ocean waters is orders of magnitude less, indicating that most of the plutonium is quite insolvable in marine waters and has been incorporated into sediments. Actinide ions in waters often are not in a state of thermodynamic equilibrium and their solubility and migration behavior is related to the form in which the nuclides were introduced into the aquatic system. Actinide solubility depends on such factors as pH(hydrolysis), E H (oxidation state), reaction with complexants (e.g. carbonate, phosphate, humic acid, etc.) sorption to surfaces of minerals and/or colloids, etc., in the water. The most significant of these variables is the oxidation sate of the metal ion. The simultaneous presence of more than one oxidation state for some actinides (e.g. plutonium) in a solution complicates actinide environmental behavior. Both Np(V)O 2 + and Pu(V)O 2 + , the most significant soluble states in natural oxic waters are relatively noncomplexing and resistant to hydrolysis and subsequent precipitation but can undergo reduction to the Pu(IV) oxidation state with its different elemental behavior. The solubility of NpO 2 + can be as high as 10 -4 M while that of PuO 2 + is more limited by reduction to the insoluble tetravalent species, Pu(OH) 4 , (pK SP - 56). The net solubility of hexavalent UO 2 2+ in sea water is also limited by hydrolysis; however, it has a relatively high concentration due to formation of carbonate complexes. The insoluble trivalent americium hydroxocarbonate, Am(CO) 3 (OH), is the limiting species for the solubility of Am(III) in sea water. Thorium is found exclusively as the tetravalent species and its solubility is limited by the formation of quite insoluble Th(OH) 4 . The chemistry of actinide ions in the environment is reviewed to show the spectrum of reactions that can occur in

Biological pathways and chemical behavior of plutonium and other actinides in the environment

International Nuclear Information System (INIS)

Dahlman, R.C.; Bondietti, E.A.; Eyman, L.D.

1976-01-01

The principal long-lived actinide elements that may enter the environment from either U or Pu fuel cycles are Pu, Am, Cm, and Np. Approximately 25% of the alpha activity estimated to be released to the atmosphere from the LMFBR fuel cycle will be contributed by 241 Am, 242 Cm, and 244 Cm. The balance of the alpha activity will come from Pu isotopes. Activities of 242 Cm, 244 Cm, 241 Am, 243 Am, and 237 Np in waste may exceed concentrations of Pu isotopes in waste after various periods of decay. Thorium and uranium isotopes may also be released by operations of the thorium fuel cycle. Environmental actinides are discussed under the following headings: sources of man-made actinide elements; pathways of exposure; environmental chemistry of actinides; uptake of actinides by plants; distribution of actinides in components of White Oak Lake; entry of actinides into terrestrial food chains; relationship between chemical behavior and uptake of actinides by organisms; and behavior of Pu in freshwater and marine food chains

Recent progress in actinide and lanthanide solvent extraction

International Nuclear Information System (INIS)

Musikas, C.; Hubert, H.; Benjelloun, N.; Vitorge, P.; Bonnin, M.; Forchioni, A.; Chachaty, C.

1983-04-01

Work in progress on actinide solvent extraction is briefly reviewed in this paper. 1 H and 31 P NMR are used to elucidate several fundamental unsolved problems concerning organophosphorous extractants often used in actinides extraction: determination of site of dialkylthiophosphate protonation and addition of basic phosphine oxide to dibutylthiophosphoric acid dimer. Extraction of Am III and Eu from high radioactivity level wastes by tetrasubsituted methylene diamides is investigated. Trivalent actinide-lanthanide group are separated by solvent extraction using soft donor ligand complexes which are more stable. The synergism of dinonylnaphtalene sulfonic acid (HDNNS) associated with several neutral donors like TBP, TOPO, amides are examined in the trivalent and tetravalent actinide extraction

Actinide removal from aqueous solution with activated magnetite

International Nuclear Information System (INIS)

Kochen, R.L.; Thomas, R.L.

1987-01-01

An actinide aqueous waste treatment process using activated magnetite has been developed at Rocky Flats. The use and effectiveness of various magnetites in lowering actinide concentrations in aqueous solution are described. Experiments indicate that magnetite particle size and pretreatment (activation of the magnetite surface with hydroxyl ions greatly influence the effective use of magnetite as an actinide adsorbent. With respect to actinide removal, Ba(OH) 2 -activated magnetite was more effective over a broader pH range than was NaOH-activated magnetite. About 50% less Ba(OH) 2 -activated magnetite was required to lower plutonium concentration from 10 -4 to 10 -8 g/l. 7 refs., 8 tabs

Reduction of minor actinides for recycling in a light water reactor

International Nuclear Information System (INIS)

Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

2015-09-01

The aim of actinide transmutation from spent nuclear fuel is the reduction in mass of high-level waste which must be stored in geological repositories and the lifetime of high-level waste; these two achievements will reduce the number of repositories needed, as well as the duration of storage. The present work is directed towards the evaluation of an advanced nuclear fuel cycle in which the minor actinides (Np, Am and Cm) could be recycled to remove most of the radioactive material; a reference of actinides production in standard nuclear fuel of uranium at the end of its burning in a BWR is first established, after a design of fuel rod containing 6% of minor actinides in a matrix of uranium from the enrichment lines is proposed, then 4 fuel rods of standard uranium are replaced by 4 actinides bars to evaluate the production and transmutation of them and finally the minor actinides reduction in the fuel is evaluated. In the development of this work the calculation tool are the codes: Intrepin-3, Casmo-4 and Simulate-3. (Author)

Actinide cation-cation complexes

International Nuclear Information System (INIS)

Stoyer, N.J.; Seaborg, G.T.

1994-12-01

The +5 oxidation state of U, Np, Pu, and Am is a linear dioxo cation (AnO 2 + ) with a formal charge of +1. These cations form complexes with a variety of other cations, including actinide cations. Other oxidation states of actinides do not form these cation-cation complexes with any cation other than AnO 2 + ; therefore, cation-cation complexes indicate something unique about AnO 2 + cations compared to actinide cations in general. The first cation-cation complex, NpO 2 + ·UO 2 2+ , was reported by Sullivan, Hindman, and Zielen in 1961. Of the four actinides that form AnO 2 + species, the cation-cation complexes of NpO 2 + have been studied most extensively while the other actinides have not. The only PuO 2 + cation-cation complexes that have been studied are with Fe 3+ and Cr 3+ and neither one has had its equilibrium constant measured. Actinides have small molar absorptivities and cation-cation complexes have small equilibrium constants; therefore, to overcome these obstacles a sensitive technique is required. Spectroscopic techniques are used most often to study cation-cation complexes. Laser-Induced Photacoustic Spectroscopy equilibrium constants for the complexes NpO 2 + ·UO 2 2+ , NpO 2 + ·Th 4+ , PuO 2 + ·UO 2 2+ , and PuO 2 + ·Th 4+ at an ionic strength of 6 M using LIPAS are 2.4 ± 0.2, 1.8 ± 0.9, 2.2 ± 1.5, and ∼0.8 M -1

Package materials, waste form

International Nuclear Information System (INIS)

Anon.

1980-01-01

The schedules for waste package development for the various host rocks were presented. The waste form subtask activities were reviewed, with the papers focusing on high-level waste, transuranic waste, and spent fuel. The following ten papers were presented: (1) Waste Package Development Approach; (2) Borosilicate Glass as a Matrix for Savannah River Plant Waste; (3) Development of Alternative High-Level Waste Forms; (4) Overview of the Transuranic Waste Management Program; (5) Assessment of the Impacts of Spent Fuel Disassembly - Alternatives on the Nuclear Waste Isolation System; (6) Reactions of Spent Fuel and Reprocessing Waste Forms with Water in the Presence of Basalt; (7) Spent Fuel Stabilizer Screening Studies; (8) Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype Canister Metals in a Simulated Wet Repository Environment; (9) Impact of Fission Gas and Volatiles on Spent Fuel During Geologic Disposal; and (10) Spent Fuel Assembly Decay Heat Measurement and Analysis

Management of actinide waste inventories in nuclear phase-out scenarios

International Nuclear Information System (INIS)

Cometto, M.; Wydler, P.; Chawla, R.

2008-01-01

The improvement of the 'radiological cleanliness' of nuclear energy is a primary goal in the development of advanced reactors and fuel cycles. The multiple recycling of actinides in advanced nuclear systems with fast neutron spectra represents a key option for reducing the potential hazard from high-level waste, especially when the fuel cycle is fully closed. Such strategies, however, involve large inventories of radiotoxic, transuranic (TRU) nuclides in the nuclear park, both in-pile and out-of-pile. The management of these inventories with the help of actinide burners is likely to become an important issue, if nuclear energy systems are eventually phased out, i.e. replaced by other types of energy systems. The present paper compares phase-out scenarios for two transmutation strategies involving fast reactors (FRs) and accelerator-driven systems (ADSs), respectively, operating in symbiosis with conventional light water reactors (LWRs). Particular objectives are to evaluate and compare the TRU reduction performance of the systems as a function of the phase-out time and to determine the appropriate phase-out length for different phase-out criteria. In this connection, an interesting aspect concerns the continuous optimisation of the fuel cycle to counterbalance the reactivity decrease due to the depletion of the fissile isotopes in the fuel. It will be shown that both FRs and ADSs can achieve the goal, provided that the phase-out operation can be continued for about a hundred years

Cerium, uranium, and plutonium behavior in glass-bonded sodalite, a ceramic nuclear waste form

International Nuclear Information System (INIS)

Lewis, M. A.; Lexa, D.; Morss, L. R.; Richmann, M. K.

1999-01-01

Glass-bonded sodalite is being developed as a ceramic waste form (CWF) to immobilize radioactive fission products, actinides, and salt residues from electrometallurgical treatment of spent nuclear reactor fuel. The CWF consists of about 75 mass % sodalite, 25 mass % glass, and small amounts of other phases. This paper presents some results and interpretation of physical measurements to characterize the CWF structure, and dissolution tests to measure the release of matrix components and radionuclides from the waste form. Tests have been carried out with specimens of the CWF that contain rare earths at concentrations similar to those expected in the waste form. Parallel tests have been carried out on specimens that have uranium or plutonium as well as the rare earths at concentrations similar to those expected in the waste forms; in these specimens UCl 3 forms UO 2 and PuCl 3 forms PuO 2 . The normalized releases of rare earths in dissolution tests were found to be much lower than those of matrix elements (B, Si, Al, Na). When there is no uranium in the CWF, the release of cerium is two to ten times lower than the release of the other rare earths. The low release of cerium may be due to its tetravalent state in uranium-free CWF. However, when there is uranium in the CWF, the release of cerium is similar to that of the other rare earths. This trivalent behavior of cerium is attributed to charge transfer or covalent interactions among cerium, uranium, and oxygen in (U,Ce)O 2

Integral migration and source term experiments on cement and bitumen waste forms

International Nuclear Information System (INIS)

Ewart, F.T.; Howse, R.M.; Sharpe, B.M.; Smith, A.J.; Thomason, H.P.; Williams, S.J.; Young, M.

1986-01-01

This is the final report of a programme of research which formed a part of the CEC joint research project into radionuclide migration in the geosphere (MIRAGE). This study addressed the aspects of integral migration and source term. The integral migration experiment simulated, in the laboratory, the intrusion of water into the repository, the leaching of radionuclides from two intermediate level wasteforms and the subsequent migration through the geosphere. The simulation consisted of a source of natural ground water which flowed over a sample of wasteform, at a controlled redox potential, and then through backfill and geological material packed in columns. The two wasteforms used here were cemented waste from the WAK plant at Karlsruhe, W. Germany and bitumenised intermediate concentrates from the Marcoule plant in France. The soluble fission products such as caesium wire rapidly released from the cemented waste but the actinides, and technetium in the reduced state, were retained in the wasteform. The release of all nuclides from the bitumenised waste was very low. (author)

Secondary Waste Cast Stone Waste Form Qualification Testing Plan

Energy Technology Data Exchange (ETDEWEB)

Westsik, Joseph H.; Serne, R. Jeffrey

2012-09-26

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). Cast Stone – a cementitious waste form, has been selected for solidification of this secondary waste stream after treatment in the ETF. The secondary-waste Cast Stone waste form must be acceptable for disposal in the IDF. This secondary waste Cast Stone waste form qualification testing plan outlines the testing of the waste form and immobilization process to demonstrate that the Cast Stone waste form can comply with the disposal requirements. Specifications for the secondary-waste Cast Stone waste form have not been established. For this testing plan, Cast Stone specifications are derived from specifications for the immobilized LAW glass in the WTP contract, the waste acceptance criteria for the IDF, and the waste acceptance criteria in the IDF Permit issued by the State of Washington. This testing plan outlines the testing needed to demonstrate that the waste form can comply with these waste form specifications and acceptance criteria. The testing program must also demonstrate that the immobilization process can be controlled to consistently provide an acceptable waste form product. This testing plan also outlines the testing needed to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support performance assessment analyses of the long-term environmental impact of the secondary-waste Cast Stone waste form in the IDF

Treatment of liquid waste containing alpha nuclides by adsorption

Energy Technology Data Exchange (ETDEWEB)

Jishu, Zeng; Xiguang, Su; Dejing, Xia; Sianhua, Fan [China Inst. of Atomic Energy, Beijing (China). Radiochemistry Dept.

1997-02-01

In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10{sup 3} Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs.

Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

Energy Technology Data Exchange (ETDEWEB)

Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

2011-09-12

The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

TSA waste stream and final waste form composition

International Nuclear Information System (INIS)

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

1993-01-01

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

A new look at actinide recycle

International Nuclear Information System (INIS)

Burch, W.D.; Croff, A.G.; Rawlins, J.A.; Schulz, W.W.

1991-01-01

This paper will address the justification for reexamination of the value of recovering the minor actinides and certain fission products from spent light-water reactor fuels and describe some of the technical progress that has been made since the major studies of a decade ago. During this time, the US Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission have begun establishing detailed criteria and regulations for geologic repositories. An examination of the hazards of waste disposal relative to the EPA release standards reveals that removal of 99.9% of the actinides (Pu, Am, and Np) reduces these hazards quite close to the EPA standards after 300 years' decay of the strontium and cesium. It may be also useful to remove and separately manage and dispose of certain of the long-lived fission products, such as 99 Tc and 129 I. Much additional work is required to fully assess the appropriate target recoveries as the hazards and risks are more closely examined and as the standards are reworked and refined. The two decades before the projected start of the US repository may present a window of opportunity to introduce several better management practices that act to simplify the repository safety issues. From a technical standpoint, significant progress has been made on recovery of the actinides from aqueous wastes though use of the TRUEX process. Additional work is required to demonstrate the application of the process to spent LWR fuels, but it appears straightforward. In addition, work at the Argonne National Laboratory on the liquid-metal reactor metal fuel cycle shows the relative simplicity of recycle of the actinides in that fast reactor cycle. Much work remains to fully demonstrate that actinides from all secondary waste streams can be removed to the target levels from both the aqueous reprocessing of LWR fuel and the pyro processes for the metal-fueled fast reactor. 9 refs., 2 figs

MIGRATION '03: 9th International Conference on Chemistry and Migration Behavior of Actinides and Fission Products in the Geosphere

International Nuclear Information System (INIS)

Kim, Geon Young; Hahn, Pil Soo; Kang, Moon Ja; Baik, Min Hoon; Kim, Seung

2003-12-01

The objectives of this report are overview of the chemistry and migration behavior of actinide for the HLW disposal safety assessment and to summarise the present status of actinide science and future developments. Actinides in HLW are very toxic and long-life time radionuclides. Therefore, the understanding of their characteristics and reaction behaviors in the deep subsurface environment is necessary for improving the reliability of HLW disposal safety assessment. This report presents an overview of the recent developments in the fundamental chemistry of actinides and fission products in natural aquifer systems, their interactions and migration in the geosphere, and the processes involved in modeling their geochemical behavior for the high level radioactive waste management. In addition, the thesis presented in MIGRATION '03 conference were described briefly. Actinide science in relation to the HLW disposal management can be classified into three main subjects; aquatic chemistry of actinides and fission products, migration behavior of radionuclides and geochemical and transport modeling. The radionuclides leached from waste forms are intruded into human environment along the groundwater flowing in the fracture around the waster disposal facility. To analyze and predict such radionuclide migration phenomena, the data that were obtained from well defined condition are required. Data obtained from studies on the chemical behaviors of actinide elements and fission products in the groundwater are essential in the safety assessment of HLW management. This report is intended to suggest the direction of R and D in actinide chemistry for the national program of HLW management in future

Transmutations of nuclear waste. Progress report RAS programme 1995: Recycling and transmutation of actinides and fission products

International Nuclear Information System (INIS)

Gruppelaar, H.; Cordfunke, E.H.P.; Konings, R.J.M.; Bultman, J.H.; Dodd, D.H.; Franken, W.M.P.; Kloosterman, J.L.; Koning, A.J.; Wichers, V.A.

1996-04-01

This report describes the progress of the Dutch RAS programme on 'Recycling and Transmutation of Actinides and Fission Products' over the year 1995, which is the second year of the 4-year programme 1994-1997. An extensive listing of reports and publications from 1991 to 1995 is given. Highlights in 1995 were: -The completion of the European Strategy Study on Nuclear Waste Transmutation as a result of which the understanding of transmutation of plutonium, minor actinides and long-lived fission products in thermal and fast reactors has been increased significantly. Important ECN contributions were given on Am, 99 Tc and 129 I transmutation options. Follow-up contracts have been obtained for the study of 100% MOX cores and accelerator-based transmutation. - Important progress in the evaluation of CANDU reactors for burning very large amounts of transuranium mixtures in inert matrices. - The first RAS irradiation experiment in the HFR, in which the transmutation of technetium and iodine was examined, has been completed and post-irradiation examination has been started. - A joint proposal of the EFTTRA cooperation for the 4 th Framework Programme of the EU, to demonstrate the feasibility of the transmutation of americium in an inert matrix by an irradiation in the HFR, has been granted. - A bilateral contract with CEA has been signed to participate in the CAPRA programme, and the work in this field has been started. - The thesis work on Actinide Transmutation in Nuclear Reactor Systems was succesfully defended. New PhD studies on Pu burning in HTGR, on nuclear data for accelerator-based systems, and on the SLM-technique for separation of actinides were started. - A review study of the use of the thorium cycle as a means for nuclear waste reduction, has been completed. A follow-up of this work is embedded in an international project for the 4th Framework Programme of the EU. (orig./DG)

Study of Thorium Phosphate Diphosphate (TPD) formation in nitric medium for the decontamination of high activity actinides bearing effluents

International Nuclear Information System (INIS)

Rousselle, Jerome

2004-01-01

Considering several activities in the nuclear industry and research, several low-level liquids wastes (LLLW) containing actinides in nitric medium must be decontaminated before being released in the environment. These liquid wastes mainly contain significant amounts of uranium(VI), neptunium(V) and plutonium(IV). In this work, two chemical ways were studied to decontaminate LLLW then to incorporate simultaneously uranium, neptunium and plutonium in the Thorium Phosphate Diphosphate (TPD). Both ways started from a nitric solution containing thorium and the actinides considered, present at their lower stable oxidation state. The first way consisted in the initial precipitation of actinide and thorium mixed oxalate. After drying the mixture containing the powder and phosphoric acid under dried argon, a poly-phase system was obtained. It was mainly composed by a thorium-actinide oxalate-phosphate. This mixture was transformed into a TPDAn solid solution (An = U, Np and/or Pu) by heating treatment at 1200 deg. C under inert atmosphere. The second way consisted in the precipitation of a precursor of TPD, identified as the Thorium Phosphate Hydrogen Phosphate loaded with the actinides considered. The gel initially formed by mixing concentrated phosphoric acid solution with the nitric actinide solution was heated at 90 - 160 deg. C in a closed PTFE container for several weeks. It led to the TPDAn solid solutions after heating at 1100 deg. C in air or under inert argon. The efficiency of both processes was evaluated through the determination of the decontamination for each actinide considered. Considering the encouraging results obtained for both kinds of processes, some complementary studies are now required before performing the effective decontamination of real Low-Level Liquid Waste using one of the methods proposed. (author) [fr

Waste management analysis for the nuclear fuel cycle: Parts I and II. Progress report, April 1--September 30, 1977

International Nuclear Information System (INIS)

Navratil, J.D.; Martella, L.L.; Smith, C.M.; Thompson, G.H.; Cash, D.L.; Childs, E.L.; Meile, L.J.

1978-01-01

A preliminary evaluation of methods for the salt waste and waste water streams and recycle preparation problems was completed. A feasibility study for removing actinides from synthetic salt waste showed that a bidentate organophosphorus extractant is the most efficient for actinide removal. The evaluation of adsorbents for removing detergents and anions from waste water suggests the use of a combination of non-ionic and a strong base ion exchange resin for best results. Evaluation of leaching and dissolution methods for the recovery of actinides from combustible waste (incinerator ash) was continued. Two promising recovery methods are: (1) reaction with cerium(IV) in nitric acid to solubilize carbon and actinide oxides, and (2) fusion with carbonate--nitrate mixtures. Silica proved to be a problem. If dissolved, it interferes with subsequent actinide recovery by forming polysilicic acid upon acidification. If not solubilized, silica-encapsulated actinide oxides may not be contacted by the dissolvent. Pretreatment of ash by refluxing with greater than or equal to 6M sodium hydroxide appears to remove silica, simplifying subsequent recovery steps

Formation of actinides in irradiated HTGR fuel elements

International Nuclear Information System (INIS)

Santos, A.M. dos.

1976-03-01

Actinide nuclide concentrations of 11 spent AVR fuel elements were determined experimentally. The burnup of the spheres varied in the range between 10% and 100% fifa, the Th : U ratio was 5 : 1. The separation procedures for actinide isolation were tested with highly irradiated ThO 2 . Separation and decontamination factors are presented. Actinide nuclide formation can be described by exponential functions of the type ln msub(nuclide) = A + B x % fifa. The empirical factors A and B were calculated performing a least squares analysis. Build-up of 232 U was discussed. According to the experimental results, 232 U is mainly produced from 230 Th, a certain amount (e.g. about 20% at a 10 5 MWd/t burnup) originated from a (n,2n) reaction of 233 U; a formation from 233 Th by a (n,2n) followed by a (n,γ) reaction was not observed. The AVR breeding rate was ascertained to be 0.5. The hazard potential of high activity waste was calculated. After a 1,000 years' storage time, the elements Pa, Am and Cm will no longer influence the total hazard index. Actinide recovery factors were proposed in order to reduce the hazard potential of the waste by an actinide removal in consideration of the reprocessing technology which is available presently. (orig.) [de

Chemical reactions in the bedrock-groundwater system of importance for the sorption of actinides

International Nuclear Information System (INIS)

Beall, G.W.; Allard, B.; Krajewski, T.; O'Kelley, G.D.

1979-01-01

Most suggested alternatives for deep underground disposal of high-level radioactive wastes rely upon several independent barriers like resistant canister materials, waste forms of low solubility in groundwaters and the use of back-fill material of low permeability and with nuclide retaining capacity. These barriers would retard the eventual release of radionuclides from the repository into the groundwater/bedrock system. The final and the only non-engineered barrier would be the host-rock itself. It would be desirable if the rock alone would be able to retain the long-lived radionuclides coming from the waste for long enough times to allow decay to harmless activity levels before they might reach the biosphere. The biological hazards from high-level reprocessng waste as well as from unreprocessed spent uranium fuel are dominated by the actinides and their daughter products (americium, plutonium, neptunium, thorium, radium) from about three hundred years after discharge from the reactor up to millions of years. In order to allow predictions of the migration of the actinides in the ground, the chemical behavior of these elements in groundwater under environmental conditions and their interactions with geologic media must be studied in detail. In this paper, studies of the sorption of americium (trivalent) and neptunium (pentavalent under oxic conditions) on some major rock-forming minerals of igneous rocks and accessory minerals are reported

Gastrointestinal absorption of actinides: a review with special reference to primate data

International Nuclear Information System (INIS)

Burkart, W.

1984-01-01

Large scale geological burial of transuranic wastes from fission power production may expose segments of future generations to trace amounts of actinides in water and food, which, via gastrointestinal absorption, could result in internal doses of alpha radiation. Gastrointestinal absorption of actinide elements is a poorly understood process. Experimental studies, primarily using rodents, often produce ambiguous results with order of magnitude fluctuations in estimates of GI absorption. Since experimental conditions like the chemical form of the fed actinides or reducing and complexing capacity of the stomach content, influence the GI transfer factor in seemingly unpredictable ways, only a better understanding of events at the molecular level will enable more reliable predictions to be made of the organ burdens resulting from actinides passing through the digestive tract. From a review of the existing literature it is apparent that in vitro research data in the area of GI uptake mechanisms (i.e. transport mediated by ion carriers in body fluids and across cell membranes) are virtually non-existant. In view of the uncertainties linked to in vivo uptake experiment, models which approximate man, i.e. derived from non-human primate studies, should be the best choice of experimental systems in which to determine reliable estimates for gastrointestinal transfer factors of actinide elements. (Auth.)

Actinide partitioning and transmutation program progress report, October 1, 1976--March 31, 1977

International Nuclear Information System (INIS)

Blomeke, J.O.; Tedder, D.W.

1977-01-01

Experimental work on the 16 tasks comprising the Actinide Partitioning and Transmutation Program was initiated at the various sites. This work included the development of conceptual material balance flowsheets which define integrated waste systems supporting an LWR fuel reprocessing plant and a mixed (U-Pu) oxide fuel refabrication plant. In addition, waste subsystems were defined for experimental evaluation. Computer analysis of partitioning-transmutation, utilizing an LMFBR for transmutation, was completed for both constant and variable waste actinide generation rates

MIGRATION '03: 9th International Conference on Chemistry and Migration Behavior of Actinides and Fission Products in the Geosphere

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon Young; Hahn, Pil Soo; Kang, Moon Ja; Baik, Min Hoon; Kim, Seung

2003-12-15

The objectives of this report are overview of the chemistry and migration behavior of actinide for the HLW disposal safety assessment and to summarise the present status of actinide science and future developments. Actinides in HLW are very toxic and long-life time radionuclides. Therefore, the understanding of their characteristics and reaction behaviors in the deep subsurface environment is necessary for improving the reliability of HLW disposal safety assessment. This report presents an overview of the recent developments in the fundamental chemistry of actinides and fission products in natural aquifer systems, their interactions and migration in the geosphere, and the processes involved in modeling their geochemical behavior for the high level radioactive waste management. In addition, the thesis presented in MIGRATION '03 conference were described briefly. Actinide science in relation to the HLW disposal management can be classified into three main subjects; aquatic chemistry of actinides and fission products, migration behavior of radionuclides and geochemical and transport modeling. The radionuclides leached from waste forms are intruded into human environment along the groundwater flowing in the fracture around the waster disposal facility. To analyze and predict such radionuclide migration phenomena, the data that were obtained from well defined condition are required. Data obtained from studies on the chemical behaviors of actinide elements and fission products in the groundwater are essential in the safety assessment of HLW management. This report is intended to suggest the direction of R and D in actinide chemistry for the national program of HLW management in future.

Room temperature ionic liquids for actinide extraction: a 'green' approach?

International Nuclear Information System (INIS)

Mohapatra, P.K.

2013-01-01

Extraction of actinides is one of the key issues in the remediation of high level radioactive wastes emanating from the back end of the nuclear fuel cycle. Effective actinide extraction makes the waste benign and ready for disposal as vitrified waste blocks in deep geological repositories. However, conventional solvent extraction methods, though being routinely used for actinide separations, have several disadvantages, which include large VOC (volatile organic compounds) inventory and generation of huge volumes of secondary wastes. Growing concern for the environment has led to the increasing interest in room temperature ionic liquids (RTIL) as an alternative to molecular diluents in myriad applications including synthesis, catalysis, separation and electrochemistry. Out of these, application of RTILs to separation science has increased enormously as can be seen from the rapid rise in the number of publications in this area in the last decade, due to their unique characteristics of high thermal stability and low volatility

Transuranic waste management program waste form development

International Nuclear Information System (INIS)

Bennett, W.S.; Crisler, L.R.

1981-01-01

To ensure that all technology necessary for long term management of transuranic (TRU) wastes is available, the Department of Energy has established the Transuranic Waste Management Program. A principal focus of the program is development of waste forms that can accommodate the very diverse TRU waste inventory and meet geologic isolation criteria. The TRU Program is following two approaches. First, decontamination processes are being developed to allow removal of sufficient surface contamination to permit management of some of the waste as low level waste. The other approach is to develop processes which will allow immobilization by encapsulation of the solids or incorporate head end processes which will make the solids compatible with more typical waste form processes. The assessment of available data indicates that dewatered concretes, synthetic basalts, and borosilicate glass waste forms appear to be viable candidates for immobilization of large fractions of the TRU waste inventory in a geologic repository

Impact of actinide recycle on nuclear fuel cycle health risks

International Nuclear Information System (INIS)

Michaels, G.E.

1992-06-01

The purpose of this background paper is to summarize what is presently known about potential impacts on the impacts on the health risk of the nuclear fuel cycle form deployment of the Advanced Liquid Metal Reactor (ALMR) 1 and Integral Fast Reactor (IF) 2 technology as an actinide burning system. In a companion paper the impact on waste repository risk is addressed in some detail. Therefore, this paper focuses on the remainder of the fuel cycle

Lithium actinide recycle process demonstration

Energy Technology Data Exchange (ETDEWEB)

Johnson, G.K.; Pierce, R.D.; McPheeters, C.C. [Argonne National Laboratory, IL (United States)

1995-10-01

Several pyrochemical processes have been developed in the Chemical Technology Division of Argonne Laboratory for recovery of actinide elements from LWR spent fuel. The lithium process was selected as the reference process from among the options. In this process the LWR oxide spent fuel is reduced by lithium at 650{degrees}C in the presence of molten LiCl. The Li{sub 2}O formed during the reduction process is soluble in the salt. The spent salt and lithium are recycled after the Li{sub 2}O is electrochemically reduced. The oxygen is liberated as CO{sub 2} at a carbon anode or oxygen at an inert anode. The reduced metal components of the LWR spent fuel are separated from the LiCL salt phase and introduced into an electrorefiner. The electrorefining step separates the uranium and transuranium (TRU) elements into two product streams. The uranium product, which comprises about 96% of the LWR spent fuel mass, may be enriched for recycle into the LWR fuel cycle, stored for future use in breeder reactors, or converted to a suitable form for disposal as waste. The TRU product can be recycled as fast reactor fuel or can be alloyed with constituents of the LWR cladding material to produce a stable waste form.

Research on Actinides in Nuclear Fuel Cycles

International Nuclear Information System (INIS)

Song, Kyu Seok; Park, Yong Joon; Cho, Young Hwan

2010-04-01

The electrochemical/spectroscopic integrated measurement system was designed and set up for spectro-electrochemical measurements of lanthanide and actinide ions in high temperature molten salt media. A compact electrochemical cell and electrode system was also developed for the minimization of reactants, and consequently minimization of radioactive waste generation. By applying these equipment, oxidation and reduction behavior of lanthanide and actinide ions in molten salt media have been made. Also, thermodynamic parameter values are determined by interpreting the results obtained from electrochemical measurements. Several lanthanide ions exhibited fluorescence properties in molten salt. Also, UV-VIS measurement provided the detailed information regarding the oxidation states of lanthanide and actinide ions in high temperature molten salt media

Research on Actinides in Nuclear Fuel Cycles

Energy Technology Data Exchange (ETDEWEB)

Song, Kyu Seok; Park, Yong Joon; Cho, Young Hwan

2010-04-15

The electrochemical/spectroscopic integrated measurement system was designed and set up for spectro-electrochemical measurements of lanthanide and actinide ions in high temperature molten salt media. A compact electrochemical cell and electrode system was also developed for the minimization of reactants, and consequently minimization of radioactive waste generation. By applying these equipment, oxidation and reduction behavior of lanthanide and actinide ions in molten salt media have been made. Also, thermodynamic parameter values are determined by interpreting the results obtained from electrochemical measurements. Several lanthanide ions exhibited fluorescence properties in molten salt. Also, UV-VIS measurement provided the detailed information regarding the oxidation states of lanthanide and actinide ions in high temperature molten salt media

Bidentate organophosphorus extractants: purification, properties and applications to removal of actinides from acidic waste solutions

International Nuclear Information System (INIS)

Schulz, W.W.; McIsaac, L.D.

1977-05-01

At both Hanford and Idaho, DHDECMP (dihexyl-N, N-diethylcarbamylmethylene phosphonate) continuous counter-current solvent extraction processes are being developed for removal of americium, plutonium, and, in some cases, other actinides from acidic wastes generated at these locations. Bench and, eventually, pilot and plant-scale testing and application of these processes have been substantially enhanced by the discovery of suitable chemical and physical methods of removing deleterious impurities from technical-grade DHDECMP. Flowsheet details, as well as various properties of purified DHDECMP extractants, are enumerated

Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

Directory of Open Access Journals (Sweden)

Nuttin A.

2012-02-01

Full Text Available The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX and uranium/plutonium mixed oxide (MOX fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

Leaching of actinides from nuclear waste glass: French experience

International Nuclear Information System (INIS)

Vernaz, E.Y.; Godon, N.

1991-01-01

The activity concentration versus time of a typical LWR glass shows that after 300 years most of the activity is attributable to three actinides (Np, Pu and Am) and to 99 Tc. This activity decreases slowly, and some 50.000 years are necessary before the activity concentration drops to the level of the richest natural ores. This paper reviews the current state of knowledge concerning the kinetics of actinide release from glass subjected to aqueous leaching

Liquid secondary waste: Waste form formulation and qualification

Energy Technology Data Exchange (ETDEWEB)

Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-31

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

Comparative waste forms study

International Nuclear Information System (INIS)

Wald, J.W.; Lokken, R.O.; Shade, J.W.; Rusin, J.M.

1980-12-01

A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings

Use of fast-spectrum reactors for actinide burning

International Nuclear Information System (INIS)

Chang, Yoon I.

1991-01-01

Finally, Integral Fast Reactor (IFR) pyroprocessing has been developed only in recent years and it appears to have potential as a relatively uncomplicated, effective actinide recovery process. In fact, actinide recycling occurs naturally in the IFR fuel cycle. Although still very much developmental, the entire IFR fuel cycle will be demonstrated on prototype-scale in conjunction with the EBR-II and its refurbished Fuel Cycle Facility starting in late 1991. A logical extension to this work, therefore, is to establish whether this IFR pyrochemical processing can be applied to extracting actinides from LWR spent fuel. This paper summarizes current thinking on the rationale for actinide recycle, its ramifications on the geologic repository and the current high-level waste management plans, and the necessary development programs. 4 figs., 4 tabs

ALMR potential for actinide consumption

International Nuclear Information System (INIS)

Cockey, C.L.; Thompson, M.L.

1992-01-01

The Advanced Liquid Metal Reactor (ALMR) is a US Department of Energy (DOE) sponsored fast reactor design based on the Power Reactor, Innovative Small Module (PRISM) concept originated by General Electric. This reactor combines a high degree of passive safety characteristics with a high level of modularity and factory fabrication to achieve attractive economics. The current reference design is a 471 MWt modular reactor fueled with ternary metal fuel. This paper discusses actinide transmutation core designs that fit the design envelope of the ALMR and utilize spent LWR fuel as startup material and for makeup. Actinide transmutation may be accomplished in the ALMR core by using either a breeding or burning configuration. Lifetime actinide mass consumption is calculated as well as changes in consumption behavior throughout the lifetime of the reactor. Impacts on system operational and safety performance are evaluated in a preliminary fashion. Waste disposal impacts are discussed. (author)

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

SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

Directory of Open Access Journals (Sweden)

Bourg Stéphane

2015-12-01

Full Text Available Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities striving to meet these challenges, carried out under the Euratom FP7 collaborative project SACSESS (Safety of Actinide Separation Processes. Emphasis is put on the safety issues of fuel reprocessing and waste storage. Two types of actinide separation processes, hydrometallurgical and pyrometallurgical, are considered, as well as related aspects of material studies, process modeling and the radiolytic stability of solvent extraction systems. Education and training of young researchers in nuclear chemistry is of particular importance for further development of this field.

Recovery of transplutonium elements from nuclear reactor waste

International Nuclear Information System (INIS)

Campbell, D.O.; Buxton, S.R.

1977-01-01

A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled. 5 claims, 2 figures

Physiochemical and spectroscopic behavior of actinides and lanthanides in solution, their sorption on minerals and their compounds formed with macromolecules

International Nuclear Information System (INIS)

Jimenez R, M.

2010-01-01

From the chemical view point, the light actinides has been those most studied; particularly the uranium, because is the primordial component of the nuclear reactors. The chemical behavior of these elements is not completely defined, since they can behave as transition metals or metals of internal transition, as they are the lanthanides. The actinides are radioactive; between them they are emitters of radiation alpha, highly toxic, of live half long and some very long, and artificial elements. For all this, to know them sometimes is preferable to use their chemical similarity with the lanthanides and to study these. In particular, the migration of emitters of radiation alpha to the environment has been studied taking as model the uranium. It is necessary to mention that actinides and lanthanides elements are in the radioactive wastes of the nuclear reactors. In the Chemistry Department of the Instituto Nacional de Investigaciones Nucleares (ININ) the researches about the actinides and lanthanides began in 1983 and, between that year and 1995 several works were published in this field. In 1993 the topic was proposed as a Department project and from then around of 13 institutional projects and managerial activity have been developed, besides 4 projects approved by the National Council of Science and Technology. The objective of the projects already developed and of the current they have been contributing knowledge for the understanding of the chemical behavior of the lanthanides and actinides, as much in solution as in the solid state, their behavior in the environment and the chemistry of their complexes with recurrent and lineal macromolecules. (Author)

Waste-form development

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1982-01-01

Contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements

Review of actinide nitride properties with focus on safety aspects

Energy Technology Data Exchange (ETDEWEB)

Albiol, Thierry [CEA Cadarache, St Paul Lez Durance Cedex (France); Arai, Yasuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

This report provides a review of the potential advantages of using actinide nitrides as fuels and/or targets for nuclear waste transmutation. Then a summary of available properties of actinide nitrides is given. Results from irradiation experiments are reviewed and safety relevant aspects of nitride fuels are discussed, including design basis accidents (transients) and severe (core disruptive) accidents. Anyway, as rather few safety studies are currently available and as many basic physical data are still missing for some actinide nitrides, complementary studies are proposed. (author)

Geochemical constraints on accumulation of actinide critical masses from stored nuclear waste in natural rock repositories. Technical report, April 1, 1978--August 31, 1978 (plus supplemental time to December 31, 1978)

International Nuclear Information System (INIS)

Brookins, D.G.

1978-01-01

Results of a literature search of abundant data on lanthanide and actinide individual and joint systematics are presented. Covered were several papers/reports about uranium solution chemistry, uranium deposits, a natural fission reactor, rare-earch deposits, manganese nodules, bedded and dome salt deposits, and miscellaneous items. This literature search is not complete but represents efforts of seven individuals attempting to gather data relevant to the objectives defined in this report. Many foreign articles, as well as many English language articles are absent. Approximately 800 articles were inspected; 69 are included in the References cited. The data search for actinides and lanthanides in natural rocks indicated that only limited segregation of the actinides U, Np, Pu, Am, and Cm from the lanthanides is possible should high-level waste be released from canisters stored in various geomedia. Supporting this were studies of Oklo and other uranium deposits, manganese nodules, monomineralic and concretion formation rates, and actinide and lathanide transport in brines. The fact that some waste canisters may, under certain conditions, contain several critical masses of one or more actinides is countered by the facts that (a) most actinides have very short half-lives and would decay before release from canisters, (b) released actinides and lanthanides, although dispersed, would be transported and deposited as a group, thus preventing point concentration of any actinides, and (c) 235 U has a much longer half-life than the other actinides, thus allowing greater time for possible reaccumulation and criticality; such a scenario would demand that 235 U be segregated effectively from other elements in the lanthanide-actinide groups.No mechanism to do this is consistent with the natural occurrences studied or the theoretical Eh-pH diagrams considered

Supercritical Carbon Dioxide-Soluble Ligands for Extracting Actinide Metal Ions from Porous Solids

International Nuclear Information System (INIS)

Joan Brennecke; Mark Dietz; Richard Barrans; Alabert Herlinger

2003-01-01

Numerous types of actinide-bearing waste materials are found throughout the DOE complex. Most of these wastes consist of large volumes of non-hazardous materials contaminated with relatively small quantities of actinide elements. Separation of these wastes into their inert and radioactive components would dramatically reduce the costs of stabilization and disposal. For example, the DOE is responsible for decontaminating concrete within 7000 surplus contaminated buildings. The best technology now available for removing surface contamination from concrete involves removing the surface layer by grit blasting, which produces a large volume of blasting residue containing a small amount of radioactive material. Disposal of this residue is expensive because of its large volume and fine particulate nature. Considerable cost savings would result from separation of the radioactive constituents and stabilization of the concrete dust. Similarly, gas diffusion plants for uranium enrichment contain valuable high-purity nickel in the form of diffusion barriers. Decontamination is complicated by the extremely fine pores in these barriers, which are not readily accessible by most cleaning techniques. A cost-effect method for the removal of radioactive contaminants would release this valuable material for salvage

Radionuclide release from high level waste forms under repository conditions in clay or granite

International Nuclear Information System (INIS)

Godon, N.; Lanza, F.

1990-01-01

The behaviour of both fully active and simulated vitrified high level waste (HLW) has been studied under conditions that are likely to occur in future repositories in clay and granite. The simulated HLW was doped with Cs, Sr, Tc and the actinides and the leaching of these elements from the glass has been measured together with their concentrations in the water of the near-field and their distribution between the various components of the repository. The diffusion coefficients of several elements in Boom clay has also been measured. The results show that the concentrations of Tc and the actinides in the near-field of a repository will be very low and that the actinides will only diffuse very slowly away from the vicinity of the glass. 24 refs., 1 figs., 7 tabs

Liquid secondary waste. Waste form formulation and qualification

Energy Technology Data Exchange (ETDEWEB)

Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-01

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testing to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford’s IDF.

Actinide separative chemistry

International Nuclear Information System (INIS)

Boullis, B.

2004-01-01

Actinide separative chemistry has focused very heavy work during the last decades. The main was nuclear spent fuel reprocessing: solvent extraction processes appeared quickly a suitable, an efficient way to recover major actinides (uranium and plutonium), and an extensive research, concerning both process chemistry and chemical engineering technologies, allowed the industrial development in this field. We can observe for about half a century a succession of Purex plants which, if based on the same initial discovery (i.e. the outstanding properties of a molecule, the famous TBP), present huge improvements at each step, for a large part due to an increased mastery of the mechanisms involved. And actinide separation should still focus R and D in the near future: there is a real, an important need for this, even if reprocessing may appear as a mature industry. We can present three main reasons for this. First, actinide recycling appear as a key-issue for future nuclear fuel cycles, both for waste management optimization and for conservation of natural resource; and the need concerns not only major actinide but also so-called minor ones, thus enlarging the scope of the investigation. Second, extraction processes are not well mastered at microscopic scale: there is a real, great lack in fundamental knowledge, useful or even necessary for process optimization (for instance, how to design the best extracting molecule, taken into account the several notifications and constraints, from selectivity to radiolytic resistivity?); and such a need for a real optimization is to be more accurate with the search of always cheaper, cleaner processes. And then, there is room too for exploratory research, on new concepts-perhaps for processing quite new fuels- which could appear attractive and justify further developments to be properly assessed: pyro-processes first, but also others, like chemistry in 'extreme' or 'unusual' conditions (supercritical solvents, sono-chemistry, could be

Solubility of actinides and surrogates in nuclear glasses

International Nuclear Information System (INIS)

Lopez, Ch.

2003-01-01

The nuclear wastes are currently incorporated in borosilicate glass matrices. The resulting glass must be perfectly homogeneous. The work discussed here is a study of actinide (thorium and plutonium) solubility in borosilicate glass, undertaken to assess the extent of actinide solubility in the glass and to understand the mechanisms controlling actinide solubilization. Glass specimens containing; actinide surrogates were used to prepare and optimize the fabrication of radioactive glass samples. These preliminary studies revealed that actinide Surrogates solubility in the glass was enhanced by controlling the processing temperature, the dissolution kinetic of the surrogate precursors, the glass composition and the oxidizing versus reducing conditions. The actinide solubility was investigated in the borosilicate glass. The evolution of thorium solubility in borosilicate glass was determined for temperatures ranging from 1200 deg C to 1400 deg C.Borosilicate glass specimens containing plutonium were fabricated. The experimental result showed that the plutonium solubility limit ranged from 1 to 2.5 wt% PuO 2 at 1200 deg C. A structural approach based on the determination of the local structure around actinides and their surrogates by EXAFS spectroscopy was used to determine their structural role in the glass and the nature of their bonding with the vitreous network. This approach revealed a correlation between the length of these bonds and the solubility of the actinides and their surrogates. (author)

Pyrometallurgical processes for recovery of actinide elements

International Nuclear Information System (INIS)

Battles, J.E.; Laidler, J.J.; McPheeters, C.C.; Miller, W.E.

1994-01-01

A metallic fuel alloy, nominally U-20-Pu-lOZr, is the key element of the Integral Fast Reactor (IFR) fuel cycle. Metallic fuel permits the use of an innovative, simple pyrometallurgical process, known as pyroprocessing, (the subject of this report), which features fused salt electrorefining of the spent fuel. Electrorefining separates the actinide elements from fission products, without producing a separate stream of plutonium. The plutonium-bearing product is contaminated with higher actinides and with a minor amount of rare earth fission products, making it diversion resistant while still suitable as a fuel material in the fast spectrum of the IFR core. The engineering-scale demonstration of this process will be conducted in the refurbished EBR-II Fuel Cycle Facility, which has entered the start-up phase. An additional pyrometallurgical process is under development for extracting transuranic (TRU) elements from Light Water Reactor (LWR) spent fuel in a form suitable for use as a feed to the IFR fuel cycle. Four candidate extraction processes have been investigated and shown to be chemically feasible. The main steps in each process are oxide reduction with calcium or lithium, regeneration of the reductant and recycle of the salt, and separation of the TRU product from the bulk uranium. Two processes, referred to as the lithium and salt transport (calcium reductant) processes, have been selected for engineering-scale demonstration, which is expected to start in late 1993. An integral part of pyroprocessing development is the treatment and packaging of high-level waste materials arising from the operations, along with the qualification of these waste forms for disposal in a geologic repository

Pyrometallurgical processes for recovery of actinide elements

Energy Technology Data Exchange (ETDEWEB)

Battles, J.E.; Laidler, J.J.; McPheeters, C.C.; Miller, W.E.

1994-01-01

A metallic fuel alloy, nominally U-20-Pu-lOZr, is the key element of the Integral Fast Reactor (IFR) fuel cycle. Metallic fuel permits the use of an innovative, simple pyrometallurgical process, known as pyroprocessing, (the subject of this report), which features fused salt electrorefining of the spent fuel. Electrorefining separates the actinide elements from fission products, without producing a separate stream of plutonium. The plutonium-bearing product is contaminated with higher actinides and with a minor amount of rare earth fission products, making it diversion resistant while still suitable as a fuel material in the fast spectrum of the IFR core. The engineering-scale demonstration of this process will be conducted in the refurbished EBR-II Fuel Cycle Facility, which has entered the start-up phase. An additional pyrometallurgical process is under development for extracting transuranic (TRU) elements from Light Water Reactor (LWR) spent fuel in a form suitable for use as a feed to the IFR fuel cycle. Four candidate extraction processes have been investigated and shown to be chemically feasible. The main steps in each process are oxide reduction with calcium or lithium, regeneration of the reductant and recycle of the salt, and separation of the TRU product from the bulk uranium. Two processes, referred to as the lithium and salt transport (calcium reductant) processes, have been selected for engineering-scale demonstration, which is expected to start in late 1993. An integral part of pyroprocessing development is the treatment and packaging of high-level waste materials arising from the operations, along with the qualification of these waste forms for disposal in a geologic repository.

Monazite-type ceramics for the immobilization of minor actinides plutonium

International Nuclear Information System (INIS)

Heuser, Julia Maria

2015-01-01

The safe disposal of radioactive waste in deep geological formations is a challenging task of present and future generations. Innovative strategies as the conditioning of radionuclides in ceramic matrices can make a contribution here. This work points out monazite-type ceramics as potential waste forms for minor actinides and Pu. Several aspects concerning nuclear disposal as well as fundamental structural information were investigated. Lanthanide phosphate endmembers (LnPO 4 ) within the stability field of monazite (Ln = La-Gd) were synthesised within the scope of this work. To extend the knowledge of monazite phases, monoclinic TbPO 4 - and DyPO 4 -phases were prepared and characterised. Tb- and Dy-phosphates are situated in the xenotime stability field close to that of monazite. They can exist as metastable monazite phases. Structural characterisations of long- and short-range order were performed by X-ray diffraction, infrared (IR) and Raman spectroscopy. Structural data could be complemented, enhanced and gaps of knowledge could be filled by the first systematic consideration of the complete Ln-monazite-series (Ln = La-Dy). Furthermore, this work focuses on Sm-monazite phases. Samarium with an atomic number of 62 is located in the middle part of the lanthanides showing the monazite structure. Accordingly, it has a mean cationic radius within the Ln-monazite-series and hence shows a relative high flexibility regarding the incorporation of radionuclides with different radii. Sintering densities of SmPO 4 ceramics were optimised by varying process parameters like pressure and number of pressing steps. An irregular texture as well as densities of 94% of the theoretical value could be achieved. The resistance of Sm-monazite against ionising radiation were examined. Radiation damages caused by the α-decay of radionuclides incorporated in a ceramic matrix were simulated by computer calculations and experimentally by heavy ion bombardment of SmPO 4 . Thin layers of

Studies of thermal-hydraulics and plant systems for actinide burning fast reactor concept

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Misumi, Masahiro; Izaki, Makoto; Koike, Hiroyuki; Tanaka, Ryokichi

1984-01-01

As one of the methods to dispose long life actinide nuclides, the actinide burning fast reactor using only actinide wastes as the fuel has been proposed. Kawasaki Heavy Industries Ltd. carried out the conceptual examination on the ABFR cooled with helium gas, cooperating with Japan Atomic Energy Research Institute, and its feasibility and problems were clarified. In this report, the setting-up of various fundamental dimensions by the parameter survey of the thermal and flowing performance of the core, the examination of the thermal and flowing characteristics of the core based on the detailed power distribution, and the examination of the plant system centering around the main cooling system are outlined. The fuel is composed of actinide oxide and diluent MgO. The diluent is used for obtaining proper excess reactivity, and MgO has been taken up also in foreign countries, considering the compatibility with actinide oxide, the easiness of reprocessing and manufacture. The fuel element is of pin type, and actinide oxide and MgO pellets are in a SUS 316 cladding tube. This ABFR can treat the wastes from ten 1000 MWe power reactors, and has the power output of about 1000 MWt. (Kako, I.)

In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Forms

International Nuclear Information System (INIS)

H.W. Stockman; S. LeStrange

2000-01-01

The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

Energy Technology Data Exchange (ETDEWEB)

H.W> Stockman; S. LeStrange

2000-09-28

The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

Mixed Waste Focus Area - Waste form initiative

International Nuclear Information System (INIS)

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

1998-01-01

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

High-performance separation and supercritical extraction of lanthanides and actinides

International Nuclear Information System (INIS)

Datta, Arpita; Sujatha, K.; Kumar, R.; Sivaraman, N.; Srinivasan, T.G.; Vasudeva Rao, P.R.

2010-01-01

Extensive studies were carried out at Chemistry Group, IGCAR for the rapid separation of individual lanthanides and actinides using dynamic ion-exchange chromatographic technique. The atom percent fission was determined from the concentrations of the lanthanide fission products, uranium and plutonium contents of dissolver solution. These advantages were exploited to significantly reduce analysis time, liquid waste generation as well as dose to operator. Supercritical fluid extraction (SFE) of actinides from waste matrices was studied in detail at our laboratory using modified supercritical carbon dioxide (Sc-CO 2 ). Complete extraction and recovery of uranium, plutonium and americium from various matrices was achieved using Sc-CO 2 modified with suitable ligands. The technique was demonstrated for the recovery of plutonium from actual waste received from different laboratories. (author)

Testing and evaluation of solidified high-level waste forms. Joint annual progress report 1983

International Nuclear Information System (INIS)

Malow, G.

1985-01-01

A second joint programme of the European Atomic Community was started in 1981 under the indirect action programme (1980-84), Action No 5 'Testing and evaluation of the properties of various potential materials for immobilizing high activity waste'. The overall objective of the research is to test various European potential solidified high-level radioactive waste forms so as to predict their behaviour after disposal. The most important aspect is to produce data to calculate the activity release from the waste products under the attack of various aqueous solutions. The experiments were partly performed under waste repository relevant conditions and partly under simplified conditions for investigating basic activity release mechanisms. The topics of the programme were: (i) studies of basic leaching mechanisms; (ii) studies of hydrothermal leaching and surface attack of waste glasses; (iii) leach test carried out in contact with granite at low water flow rates; (iv) static leach tests with specimen surrounded by canister and backfill materials; (v) specific isotope leach tests in slowly flowing water; (vi) leach test of actinide spiked samples; (vii) leach tests of highly radioactive samples; (viii) leach tests of alpha radiation stability; (ix) studies of mechanical stability; (x) studies of mineral phases as model compounds and phase relations

Enhancement of cemented waste forms by supercritical CO2 carbonation of standard portland cements

International Nuclear Information System (INIS)

Rubin, J.B.; Carey, J.; Taylor, C.M.V.

1997-01-01

We are conducting experiments on an innovative transformation concept, using a traditional immobilization technique, that may significantly reduce the volume of hazardous or radioactive waste requiring transport and long-term storage. The standard practice for the stabilization of radioactive salts and residues is to mix them with cements, which may include additives to enhance immobilization. Many of these wastes do not qualify for underground disposition, however, because they do not meet disposal requirements for free liquids, decay heat, head-space gas analysis, and/or leachability. The treatment method alters the bulk properties of a cemented waste form by greatly accelerating the natural cement-aging reactions, producing a chemically stable form having reduced free liquids, as well as reduced porosity, permeability and pH. These structural and chemical changes should allow for greater actinide loading, as well as the reduced mobility of the anions, cations, and radionuclides in aboveground and underground repositories. Simultaneously, the treatment process removes a majority of the hydrogenous material from the cement. The treatment method allows for on-line process monitoring of leachates and can be transported into the field. We will describe the general features of supercritical fluids, as well as the application of these fluids to the treatment of solid and semi-solid waste forms. some of the issues concerning the economic feasibility of industrial scale-up will be addressed, with particular attention to the engineering requirements for the establishment of on-site processing facilities. Finally, the initial results of physical property measurements made on portland cements before and after supercritical fluid processing will be presented

Supercritical fluid carbon dioxide extraction of actinides

International Nuclear Information System (INIS)

Rao, Ankita; Tomar, B.S.

2016-01-01

Supercritical fluid extraction (SFE) is a process akin to liquid-liquid or solvent extraction where a Supercritical fluid (SCF) is contacted with a solid/ liquid matrix for the purpose of separating the component of interest from the original matrix. Carbon dioxide is a preferred choice as supercritical fluid (SCF) owing to its moderate critical parameter (P c = 7.38 MPa and T c = 304.1K) coupled with radiation and chemical stability, non toxic nature and low cost. Despite widespread applications for extraction of organic compounds and associated advantages especially liquid waste minimization, the SFE of metal ions was left unexplored for quite some time, as direct metal ion extraction is inefficient due charge neutralization requirement and weak solute-solvent interaction. Neutral SCF soluble metal-ligand complexation is imperative and SFE of actinides was reported only in 1994. Several studies have been carried out on SFE of uranium, thorium and plutonium from nitric acid medium employing different sets of ligands (organophosphorus, diketones, amides). Especially attractive is the possibility of direct dissolution and extraction of actinides employing ligand-acid adducts (like TBP.HNO 3 adduct) from solid matrices of different stages of nuclear fuel cycle viz. ores, spent nuclear fuels and radioactive wastes. Also, partitioning of actinides from fission products has been explored in spent nuclear fuel. These studies on supercritical fluid extraction of actinides indicate a more efficient and environmentally sustainable technology. (author)

Feasibility studies of actinide recycle in LMFBRs as a waste management alternative

International Nuclear Information System (INIS)

Beaman, S.L.; Aitken, E.A.

1976-01-01

Actinide recycle in LMFBRs offers an attractive alternative on long-term storage of the actinides. The concept will not significantly affect the performance of the LMFBR, but will affect other parts of the nuclear fuel cycle. Assuming that hands-on maintenance will be allowed for Pu-recycle fuel fabrication facilities, the transplutonium actinides should be kept separate from the PuO 2 --UO 2 fuel. Thus, the ''reference'' recycle scheme should be defined as a scheme in which the actinides are recycled in target assemblies. The target assemblies should be reprocessed either in batches separate from spent-fuel batches or in a separate, relatively small, special purpose reprocessing plant. The target assemblies should be fabricated in a special purpose, remotely maintained facility

PRODUCTION OF ACTINIDE METAL

Science.gov (United States)

Knighton, J.B.

1963-11-01

A process of reducing actinide oxide to the metal with magnesium-zinc alloy in a flux of 5 mole% of magnesium fluoride and 95 mole% of magnesium chloride plus lithium, sodium, potassium, calcium, strontium, or barium chloride is presented. The flux contains at least 14 mole% of magnesium cation at 600-- 900 deg C in air. The formed magnesium-zinc-actinide alloy is separated from the magnesium-oxide-containing flux. (AEC)

Waste form development

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1982-01-01

In this program, contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements (both as they exist and as they are modified with time). 6 tables

Structural characterization of the Actinides (III) and (IV) - DOTA complexes

International Nuclear Information System (INIS)

Audras, Matthieu

2014-01-01

The polyamino-carboxylate anions have been identified as compounds of interest in the operations of actinide separation, in actinide migration in the environment and in human radio-toxicology. The structural characterization of complexes formed between actinides and polyamino-carboxylates ligands is essential for a better understanding of actinide-ligands interactions. Among the polyamino-carboxylate anions, the DOTA ligand (1,4,7,10-tetraaza-cyclododecane tetraacetic acid) is described as a very strong complexing agent of the lanthanides(III), but has been little studied with actinides. The objective of this thesis is to describe the complexes formed between the actinides (III) and (IV) and the DOTA ligand, and compare them with the lanthanide complexes. For this, an approach has been introduced to characterize the complexes by complementary analytical techniques (spectrophotometry, electro-spray ionization mass spectrometry, NMR, EXAFS, electrochemistry), but also by calculations of theoretical chemistry to help the interpretation of the experimental data. The formation of a 1:1 complex is observed with the actinides(III) (plutonium and americium) as for lanthanides(III): rapid formation of intermediate species which evolves slowly towards the formation of a limit complex. Within this complex, the cation is located inside the cavity formed by the ligand. Four nitrogen atoms and four oxygen atoms from the carboxylate functions are involved in the coordination sphere of the cation. However, differences were observed in the bond lengths formed between the cation and the nitrogen atoms (the bonds are somewhat shorter in the case of actinide complexes) as well as the complexation kinetics, which is slightly faster for the actinides(III) than for lanthanide(III) ions of equivalent radius. The same behavior was observed in solution upon complexation of actinides(IV) (uranium, plutonium and neptunium): slow formation of a 1:1 complex (actinide(IV):ligand) in wherein the

Geochemistry of actinides. Application to the storage of high level radioactive wastes. Under the supervision of Mr Michel Treuil

International Nuclear Information System (INIS)

Bouabdallah, Noureddine; Cunault, Jean-Baptiste; Houtin, Gwenaelle; Leborgne, Francois; Lemaire, Celine; Lemarchand, Damien; Quitte, Ghylaine

1997-06-01

This collective research report first addresses the chemistry of actinides with a description of their atomic orbitals and the study of their behaviour in solution. The author addresses several aspects: historical overview on actinides, radioactivity, chemical reactions in aqueous solution, redox chemistry, speciation in solution with respect to water characteristics in deep storage conditions. The second part gathers several studies performed on a natural laboratory (the Oklo site in which nuclear reactions occurred about 2 billions years ago) and reports the modelling of radionuclide transfer within a geological system (the model is applied to the Oklo site). The third part addresses issues related to the nuclear fuel cycle, and the storage modes and materials envisaged and involved regarding the storage of high level radioactive wastes, notably in France

Minor actinides transmutation scenario studies with PWRs, FRs and moderated targets

International Nuclear Information System (INIS)

Grouiller, J.P.; Pillon, S.; Saint Jean, C. de; Varaine, F.; Leyval, L.; Vambenepe, G.; Carlier, B.

2003-01-01

Using current technologies, we have demonstrated in this study that it is theoretically possible to obtain different minor actinide transmutation scenarios with a significant gain on the waste radiotoxicity inventory. The handling of objects with Am+Cm entails the significant increase of penetrating radiation sources (neutron and γ) whatever mixed scenario is envisioned; the PWR and FR scenario involving the recycling of Am + Cm in the form of targets results in the lowest flow. In the light of these outcomes, the detailed studies has allowed to design a target sub assembly with a high fission rate (90%) and define a drawing up of reprocessing diagram with the plant head, the minor actinide separation processes (PUREX, DIAMEX and SANEX). Some technological difficulties appear in manipulating curium, principally in manufacturing where the wet process ('sol-gel') is not acquired for (Am+Cm). (author)

Subject bibliography of radioactive waste management publications at Pacific Northwest Laboratory, 1975-1978

International Nuclear Information System (INIS)

Powell, J.A.

1981-10-01

This bibliography contains publications from 1975 to 1978 written by PNL staff. PNL translations are also announced in this document. The following areas are covered: actinides; airborne wastes; alternative waste forms; calcination; characterization; containers; decontamination; disposal; high-level wastes; liquid wastes; radionuclide migration; safety; separation processes; soils; solidification; storage; transport; transuranic waste; and vitrification

Minor actinide transmutation on PWR burnable poison rods

International Nuclear Information System (INIS)

Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

2015-01-01

Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

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

International Nuclear Information System (INIS)

Mitchell, D.E.

1986-01-01

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

Waste acceptance product specifications for vitrified high-level waste forms

International Nuclear Information System (INIS)

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

1993-01-01

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

Actinide occurrences in sediments following ground disposal of acid wastes at 216-Z-9

International Nuclear Information System (INIS)

Ames, L.L.

1976-01-01

Liquid acid wastes from a Pu recovery facility at Hanford were released to the ground via structures collectively termed trenches from 1955 through 1962. Data are presented from a study of the microdistribution of Am and Pu in samples from the 216-Z-9 trench. Solution sediment relationships and associated actinide removal mechanisms under acid conditions were studied. Core wells were drilled into the sediments in which this covered trench is located and in the immediate vicinity to obtain samples for quantitative mineralogical analysis and comparison of sediments from various depths of contaminated and noncontaminated areas. Analytical techniques are described and results are reported

Preliminary Hanford Waste Vitrification Plan Waste Form Qualification Plan

International Nuclear Information System (INIS)

Nelson, J.L.

1987-09-01

This Waste Form Qualification Plan describes the waste form qualification activities that will be followed during the design and operation of the Hanford Waste Vitrification Plant to ensure that the vitrified Hanford defense high-level wastes will meet the acceptance requirements of the candidate geologic repositories for nuclear waste. This plan is based on the defense waste processing facility requirements. The content of this plan is based on the assumption that the Hanford Waste Vitrification Plant high-level waste form will be disposed of in one of the geologic repository projects. Proposed legislation currently under consideration by Congress may change or delay the repository site selection process. The impacts of this change will be assessed as details of the new legislation become available. The Plan describes activities, schedules, and programmatic interfaces. The Waste Form Qualification Plan is updated regularly to incorporate Hanford Waste Vitrification Plant-specific waste acceptance requirements and to serve as a controlled baseline plan from which changes in related programs can be incorporated. 10 refs., 5 figs., 5 tabs

The chemistry of the actinide elements. Volume I

International Nuclear Information System (INIS)

Katz, J.J.; Seaborg, G.T.; Morss, L.R.

1986-01-01

The Chemistry of the Actinide Elements is a comprehensive, contemporary and authoritative exposition of the chemistry and related properties of the 5f series of elements: actinium, thorium, protactinium, uranium and the first eleven. This second edition has been completely restructured and rewritten to incorporate current research in all areas of actinide chemistry and chemical physics. The descriptions of each element include accounts of their history, separation, metallurgy, solid-state chemistry, solution chemistry, thermo-dynamics and kinetics. Additionally, separate chapters on spectroscopy, magnetochemistry, thermodynamics, solids, the metallic state, complex ions and organometallic compounds emphasize the comparative chemistry and unique properties of the actinide series of elements. Comprehensive lists of properties of all actinide compounds and ions in solution are given, and there are special sections on such topics as biochemistry, superconductivity, radioisotope safety, and waste management, as well as discussion of the transactinides and future elements

Thermochemical and thermophysical properties of minor actinide compounds

International Nuclear Information System (INIS)

Minato, Kazuo; Takano, Masahide; Otobe, Haruyoshi; Nishi, Tsuyoshi; Akabori, Mitsuo; Arai, Yasuo

2009-01-01

Burning or transmutation of minor actinides (MA: Np, Am, Cm) that are classified as the high-level radioactive waste in the current nuclear fuel cycle is an option for the advanced nuclear fuel cycle. Although the thermochemical and thermophysical properties of minor actinide compounds are essential for the design of MA-bearing fuels and analysis of their behavior, the experimental data on minor actinide compounds are limited. To support the research and development of the MA-bearing fuels, the property measurements were carried out on minor actinide nitrides and oxides. The lattice parameters and their thermal expansions were measured by high-temperature X-ray diffractometry. The specific heat capacities were measured by drop calorimetry and the thermal diffusivities by laser-flash method. The thermal conductivities were determined by the specific heat capacities, thermal diffusivities and densities. The oxygen potentials were measured by electromotive force method.

General survey of applications which require actinide nuclear data

International Nuclear Information System (INIS)

Raman, S.

1976-01-01

This review paper discusses the actinide waste problem, the buildup of toxic isotopes in the fuel, the neutron activity associated with irradiated fuel, the 252 Cf buildup problem, and the production of radioisotope power sources as broad areas that require actinide cross-section data. Decay data enter into the area of radiological safety and health physics. This paper also discusses a few cross-section measurements in progress at the Oak Ridge Electron Linear Accelerator. The availability of actinide samples through the Transuranium Program at Oak Ridge is discussed in considerable detail. The present data status with respect to the various applications is reviewed along with recommendations for improving the data base

Status report on actinide and fission product transmutation studies

International Nuclear Information System (INIS)

1997-06-01

The management of radioactive waste is one of the key issues in today's political and public discussions on nuclear energy. One of the fields that looks into the future possibilities of nuclear technology is the neutronic transmutation of actinides and of some most important fission products. Studies on transmutation of actinides are carried out in various countries and at an international level. This status report which gives an up-to-date general overview of current and planned research on transmutation of actinides and fission products in non-OECD countries, has been prepared by a Technical Committee meeting organized by the IAEA in September 1995. 168 refs, 16 figs, 34 tabs

European Europart integrated project on actinide partitioning

International Nuclear Information System (INIS)

Madic, C.; Hudson, M.J.

2005-01-01

This poster presents the objectives of EUROPART, a scientific integrated project between 24 European partners, mostly funded by the European Community within the FP6. EUROPART aims at developing chemical partitioning processes for the so-called minor actinides (MA) contained in nuclear wastes, i.e. from Am to Cf. In the case of dedicated spent fuels or targets, the actinides to be separated also include U, Pu and Np. The techniques considered for the separation of these radionuclides belong to the fields of hydrometallurgy and pyrometallurgy, as in the previous FP5 programs named PARTNEW and PYROREP. The two main axes of research within EUROPART will be: The partitioning of MA (from Am to Cf) from high burn-up UO x fuels and multi-recycled MOx fuels; the partitioning of the whole actinide family for recycling, as an option for advanced dedicated fuel cycles (and in connection with the studies to be performed in the EUROTRANS integrated project). In hydrometallurgy, the research is organised into five Work Packages (WP). Four WP are dedicated to the study of partitioning methods mainly based on the use of solvent extraction methods, one WP is dedicated to the development of actinide co-conversion methods for fuel or target preparation. The research in pyrometallurgy is organized into four WP, listed hereafter: development of actinide partitioning methods, study of the basic chemistry of trans-curium elements in molten salts, study of the conditioning of the wastes, some system studies. Moreover, a strong management team will be concerned not only with the technical and financial issues arising from EUROPART, but also with information, communication and benefits for Europe. Training and education of young researchers will also pertain to the project. EUROPART has also established collaboration with US DOE and Japanese CRIEPI. (authors)

ACTINET: a European Network for Actinide Sciences

International Nuclear Information System (INIS)

Bernard Boullis; Pascal Chaix

2006-01-01

Full text of publication follows: The research in Actinide sciences appear as a strategic issue for the future of nuclear systems. Sustainability issues are clearly in connection with the way actinide elements are managed (either addressing saving natural resource, or decreasing the radiotoxicity of the waste). The recent developments in the field of minor actinide P and T offer convincing indications of what could be possible options, possible future processes for the selective recovery of minor actinides. But they point out, too, some lacks in the basic understanding of key-issues (such as for instance the control An versus Ln selectivity, or solvation phenomena in organic phases). Such lacks could be real obstacles for an optimization of future processes, with new fuel compounds and facing new recycling strategies. This is why a large and sustainable work appears necessary, here in the field of basic actinide separative chemistry. And similar examples could be taken from other aspects of An science, for various applications (nuclear fuel or transmutation targets design, or migration issues,): future developments need a strong, enlarged, scientific basis. The Network ACTINET, established with the support of the European Commission, has the following objectives: - significantly improve the accessibility of the major actinide facilities to the European scientific community, and form a set of pooled facilities, as the corner-stone of a progressive integration process, - improve mobility between the member organisations, in particular between Academic Institutions and National Laboratories holding the pooled facilities, - merge part of the research programs conducted by the member institutions, and optimise the research programs and infrastructure policy via joint management procedures, - strengthen European excellence through a selection process of joint proposals, and reduce the fragmentation of the community by putting critical mass of resources and expertise on

Moessbauer effect studies with actinides

International Nuclear Information System (INIS)

Stone, J.A.

1966-01-01

Moessbauer resonance studies in the actinide elements offer a new technique for measuring solid-state properties to a region of the periodic chart where such information is relatively sparse. It is well known that the actinides, the elements with atomic numbers from 90 to 103, form a transition series due to filling of the 5f electron shell, analogous to the rare-earth series in which the 4f shell is filled. Like the rare earths, the actinide metals and compounds are expected to exhibit a variety of interesting magnetic properties, but, unlike the rare earths, there have been few studies of the magnetic behaviour of actinides, and these properties are largely unknown. The chemical properties of the actinides have been studied somewhat more extensively, and, in contrast to the rare earths, form a multiplicity of stable valence states, especially in the lighter members of the series. It is just these properties, magnetic and chemical, for which the Moessbauer effect is a valuable probe, sensitive to the magnetic and electric environment of an atom. The rare-earth series has been a particularly fruitful region in terms of the number of elements which have been shown to exhibit the Moessbauer effect, and for this reason the exploitation of the Moessbauer effect to yield new solid-state and chemical information on the rare earths is a highly active field of research today. There is every reason to believe that the actinides can be similarly studied by the Moessbauer effect. 43 refs, 6 figs, 4 tabs

Application of plasma shield technology to the reduction, treatment, and disposal of hazardous organic and/or mixed wastes with actinide recovery

International Nuclear Information System (INIS)

Adams, B.T.; Vaughan, L.L.; Joyce, E.L. Jr.; Bieniewski, T.M.

1990-01-01

Los Alamos research activities are currently directed at the application of the shielded hydrogen plasma torch to the direct production of actinide metals from a UF 6 feedstock. Two broad classes of thermal plasma reactors are currently in widespread use: the direct current (dc) arc jet system and the radio frequency (rf) inductively coupled system. Los Alamos has improved upon the basic rf plasma tube design using the concept of a transformer. The unique feature of the Los Alamos tube is a segmented, cooled, internal radiation shield. The Los Alamos shielded plasma torch routinely achieves temperatures exceeding 10,000 K and electron densities of 10 16 /cm 3 when operated continuously at one atmosphere of argon. These highly energetic conditions are sufficient to dissociate most chemical compounds into their constituent atoms. Based upon these characteristics, Los Alamos is currently investigating the application of the shielded plasma torch technology to the destruction of organic and mixed hazardous wastes, as well as the direct production of actinide metals from the halides and oxides, without the cogeneration of contaminated wastes. 5 refs., 4 figs

Position of actinide elements in the periodic table : evolution of a generalised form (Preprint no. SSC-33)

International Nuclear Information System (INIS)

Nathaniel, T.N.

1991-01-01

All the actinides and lanthanides are placed in III group and are shown separately, in the most popular forms of the periodic table, including the latest version of IUPAC. This has been leading to an avoidable debate i.e. whether to place La and Ac or Lu and Lw below Sc and Y in the III-B group. Apart from this, the group characteristic nature of most of the actinides elements, demands that a more appropriate position is to be given to these f block elements. The situation is more complex in the higher periods, in which new elements belonging to g block, h block, ... are to be included. Also, there seems to be no consensus in following a uniform and acceptable subgroup notation. The author arrived at a new form of the periodic table to successfully sort out all these issues. The table is presented. (author). 10 refs., 1 fig

Quantification of the Partitioning Ratio of Minor Actinide Surrogates between Zirconolite and Glass in Glass-Ceramic for Nuclear Waste Disposal.

Science.gov (United States)

Liao, Chang-Zhong; Liu, Chengshuai; Su, Minhua; Shih, Kaimin

2017-08-21

Zirconolite-based glass-ceramic is considered a promising wasteform for conditioning minor actinide-rich nuclear wastes. Recent studies on this wasteform have sought to enhance the partitioning ratio (PR) of minor actinides in zirconolite crystal. To optimize the PR in the SiO 2 -Al 2 O 3 -CaO-TiO 2 -ZrO 2 system, a novel conceptual approach, which can be derived from the chemical composition and quantity of zirconolite crystal in glass-ceramic, was introduced based on the results of Rietveld quantitative X-ray diffraction analysis and transmission electron microscopy energy dispersive X-ray spectroscopy. To verify this new conceptual approach, the influences of the crystallization temperature, the concentration of additives, and ionic radii on the PR of various surrogates (Ce, Nd, Gd, and Yb) in zirconolite were examined. The results reveal that the PR of Nd 3+ in zirconolite can be as high as 41%, but it decreases as the crystallization temperature increases. The quantities of all phases (including crystalline and amorphous) remained nearly constant when increasing the loading of Nd 2 O 3 in glass-ceramic products crystallized at 1050 °C for 2 h. Correspondingly, the PR of Nd 3+ decreases in a linear fashion with the loading contents of Nd 2 O 3 . The radius of ions also has a great influence on the PR, and an increase in the ionic radius leads to a decrease in the PR. This new approach will be an important tool to facilitate the exploration of a glass-ceramic matrix for the disposal of minor actinide-rich nuclear wastes.

Mechanical environmental transport of actinides and ¹³⁷Cs from an arid radioactive waste disposal site.

Science.gov (United States)

Snow, Mathew S; Clark, Sue B; Morrison, Samuel S; Watrous, Matthew G; Olson, John E; Snyder, Darin C

2015-10-01

Aeolian and pluvial processes represent important mechanisms for the movement of actinides and fission products at the Earth's surface. Soil samples taken in the early 1970's near a Department of Energy radioactive waste disposal site (the Subsurface Disposal Area, SDA, located in southeastern Idaho) provide a case study for studying the mechanisms and characteristics of environmental actinide and (137)Cs transport in an arid environment. Multi-component mixing models suggest actinide contamination within 2.5 km of the SDA can be described by mixing between 2 distinct SDA end members and regional nuclear weapons fallout. The absence of chemical fractionation between (241)Am and (239+240)Pu with depth for samples beyond the northeastern corner and lack of (241)Am in-growth over time (due to (241)Pu decay) suggest mechanical transport and mixing of discrete contaminated particles under arid conditions. Occasional samples northeast of the SDA (the direction of the prevailing winds) contain anomalously high concentrations of Pu with (240)Pu/(239)Pu isotopic ratios statistically identical to those in the northeastern corner. Taken together, these data suggest flooding resulted in mechanical transport of contaminated particles into the area between the SDA and a flood containment dike in the northeastern corner, following which subsequent contamination spreading in the northeastern direction resulted from wind transport of discrete particles. Copyright © 2015 Elsevier Ltd. All rights reserved.

The INE-Beamline for actinide science at ANKA

Science.gov (United States)

Rothe, J.; Butorin, S.; Dardenne, K.; Denecke, M. A.; Kienzler, B.; Löble, M.; Metz, V.; Seibert, A.; Steppert, M.; Vitova, T.; Walther, C.; Geckeis, H.

2012-04-01

Since its inauguration in 2005, the INE-Beamline for actinide research at the synchrotron source ANKA (KIT North Campus) provides dedicated instrumentation for x-ray spectroscopic characterization of actinide samples and other radioactive materials. R&D work at the beamline focuses on various aspects of nuclear waste disposal within INE's mission to provide the scientific basis for assessing long-term safety of a final nuclear waste repository. The INE-Beamline is accessible for the actinide and radiochemistry community through the ANKA proposal system and the European Union Integrated Infrastructure Initiative ACTINET-I3. Experiments with activities up to 1 × 10+6 times the European exemption limit are feasible within a safe but flexible containment concept. Measurements with monochromatic radiation are performed at photon energies varying between ˜2.1 keV (P K-edge) and ˜25 keV (Pd K-edge), including the lanthanide L-edges and the actinide M- and L3-edges up to Cf. The close proximity of the INE-Beamline to INE controlled area labs offers infrastructure unique in Europe for the spectroscopic and microscopic characterization of actinide samples. The modular beamline design enables sufficient flexibility to adapt sample environments and detection systems to many scientific questions. The well-established bulk techniques x-ray absorption fine structure (XAFS) spectroscopy in transmission and fluorescence mode have been augmented by advanced methods using a microfocused beam, including (confocal) XAFS/x-ray fluorescence detection and a combination of (micro-)XAFS and (micro-)x-ray diffraction. Additional instrumentation for high energy-resolution x-ray emission spectroscopy has been successfully developed and tested.

The INE-Beamline for actinide science at ANKA

International Nuclear Information System (INIS)

Rothe, J.; Dardenne, K.; Denecke, M. A.; Kienzler, B.; Loeble, M.; Metz, V.; Steppert, M.; Vitova, T.; Geckeis, H.; Butorin, S.; Seibert, A.; Walther, C.

2012-01-01

Since its inauguration in 2005, the INE-Beamline for actinide research at the synchrotron source ANKA (KIT North Campus) provides dedicated instrumentation for x-ray spectroscopic characterization of actinide samples and other radioactive materials. R and D work at the beamline focuses on various aspects of nuclear waste disposal within INE's mission to provide the scientific basis for assessing long-term safety of a final nuclear waste repository. The INE-Beamline is accessible for the actinide and radiochemistry community through the ANKA proposal system and the European Union Integrated Infrastructure Initiative ACTINET-I3. Experiments with activities up to 1 x 10 +6 times the European exemption limit are feasible within a safe but flexible containment concept. Measurements with monochromatic radiation are performed at photon energies varying between ∼2.1 keV (P K-edge) and ∼25 keV (Pd K-edge), including the lanthanide L-edges and the actinide M- and L3-edges up to Cf. The close proximity of the INE-Beamline to INE controlled area labs offers infrastructure unique in Europe for the spectroscopic and microscopic characterization of actinide samples. The modular beamline design enables sufficient flexibility to adapt sample environments and detection systems to many scientific questions. The well-established bulk techniques x-ray absorption fine structure (XAFS) spectroscopy in transmission and fluorescence mode have been augmented by advanced methods using a microfocused beam, including (confocal) XAFS/x-ray fluorescence detection and a combination of (micro-)XAFS and (micro-)x-ray diffraction. Additional instrumentation for high energy-resolution x-ray emission spectroscopy has been successfully developed and tested.

Assessment of Partitioning Processes for Transmutation of Actinides

International Nuclear Information System (INIS)

2010-04-01

To obtain public acceptance of future nuclear fuel cycle technology, new and innovative concepts must overcome the present concerns with respect to both environmental compliance and proliferation of fissile materials. Both these concerns can be addressed through the multiple recycling of all transuranic elements (TRUs) in fast neutron reactor. This is only possible through a process known as partitioning and transmutation scheme (P and T) as this scheme is expected to reduce the long term radio-toxicity as well as the radiogenic heat production of the nuclear waste. Proliferation resistance of separated plutonium could further be enhanced by mixing with self-generated minor actinides. In addition, P and T scheme is expected to extend the nuclear fuel resources on earth about 100 times because of the recycle and reuse of fissile actinides. Several Member States are actively pursuing the research in the field of P and T and consequently several IAEA publications have addressed this topic. The present coordinated research project (CRP) focuses on the potentials in minimizing the residual TRU inventories of the discharged nuclear waste and in enhancing the proliferation resistance of the future civil nuclear fuel cycle. Partitioning approaches can be grouped into aqueous- (hydrometallurgical) and pyroprocesses. Several aqueous processes based on sequential separation of actinides from spent nuclear fuel have been developed and tested at pilot plant scale. In view of the proliferation resistance of the intermediate and final products of a P and T scheme, a group separation of all actinides together is preferable. The present CRP has gathered experts from different organisations and institutes actively involved in developing P and T scheme as mentioned in the list of contributors and also taken into consideration the studies underway in France and the UK. The scientific objectives of the CRP are: To minimize the environmental impact of actinides in the waste stream; To

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

Nuclear fuel cycle-oriented actinides separation in China

Energy Technology Data Exchange (ETDEWEB)

Chen, Jing; He, Xihong; Wang, Jianchen [Tsinghua Univ., Beijing (China). Inst. of Nuclear and New Energy Technology

2014-04-01

In the last decades, the separation of actinides was widely and continuously studied in China. A few kinds of salt-free reductants to adjust Pu and Np valences have been investigated. N,N-dimethylhydroxylamine is a good reductant with high reduction rate constants for the co-reduction of Pu(IV) and Np(VI), and monomethylhydrazine is a simple compound for the individual reduction of Np(VI). Advanced PUREX based on Organic Reductants (APOR) was proposed. Trialkylphosphine oxide (TRPO) with a single functional group was found to possess strong affinity to tri-, tetra- and hexa-valent actinides. TRPO process has been first explored in China for actinides partitioning from high level waste and the good partitioning performance was demonstrated by the hot test. High extraction selectivity for trivalent actinides over lanthanides by dialkyldithiophosphinic acids was originally found in China. A separation process based on purified Cyanex 301 for the separation of Am from lanthanides was presented and successfully tested in a battery of miniature centrifugal contactors. (orig.)

Development of quantitative analytical methods for the control of actinides in a pyrochemical partitioning process

International Nuclear Information System (INIS)

Abousahl, S.; Belle, P. van; Eberle, H.; Ottmar, H.; Lynch, B.; Vallet, P.; Mayer, K.; Ougier, M.

2005-01-01

Advanced nuclear fuel cycles are being developed in order to reduce the long-term radiotoxicity of highly radioactive waste. Pyrochemical partitioning techniques appear particularly attractive for advanced fuel cycles in which the minor actinides are recycled. The electrochemical processes of practical importance are the electrorefining process and the liquid-liquid extraction of transuranic (TRU) elements from fission products using either non-miscible molten metal or molten salt-metal phases. Analytical methods for the accurate assay of actinide elements in these matrices needed to be developed. A quantitative assay is required in order to establish a material balance for process development and - at a later stage - for accountancy and control purposes. To this end radiometric techniques such as energy-dispersive X-ray fluorescence analysis (XRF), neutron coincidence counting (NCC) and high-resolution gamma spectrometry (HRGS) were extensively employed for the quantitative determination of actinides (U, Np, Pu, Am, Cm) in process samples. Comparative analyses were performed using inductively coupled plasma mass spectrometry (ICP-MS). The respective samples were available in small quantities (≅ 100 mg) either in the form of eutectic salt or in metallic form with Cd, Zr or Bi as major metallic matrix constituents. (orig.)

Radionuclide Retention in Concrete Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Mattigod, Shas V.; Bovaird, Chase C.; Wellman, Dawn M.; Wood, Marcus I.

2010-09-30

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. The information presented in the report provides data that 1) quantify radionuclide retention within concrete waste form materials similar to those used to encapsulate waste in the Low-Level Waste Burial Grounds (LLBG); 2) measure the effect of concrete waste form properties likely to influence radionuclide migration; and 3) quantify the stability of uranium-bearing solid phases of limited solubility in concrete.

The effect of corrosion product colloids on actinide transport

International Nuclear Information System (INIS)

Gardiner, M.P.; Smith, A.J.; Williams, S.J.

1992-01-01

The near field of the proposed UK repository for ILW/LLW will contain containers of conditioned waste in contact with a cementious backfill. It will contain significant quantities of iron and steel, Magnox and Zircaloy. Colloids deriving from their corrosion products may possess significant sorption capacity for radioelements. If the colloids are mobile in the groundwater flow, they could act as a significant vector for activity transport into the far field. The desorption of plutonium and americium from colloidal corrosion products of iron and zirconium has been studied under chemical conditions representing the transition from the near field to the far field. Desorption R d values of ≥ 5 x 10 6 ml g -1 were measured for both actinides on these oxides and hydroxides when actinide sorption took place under the near-field conditions and desorption took place under the far-field conditions. Desorption of the actinides occurred slowly from the colloids under far-field conditions when the colloids had low loadings of actinide and more quickly at high loadings of actinide. Desorbed actinide was lost to the walls of the experimental vessel. (author)

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Science.gov (United States)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

Preliminary flowsheet for the conversion of Hanford high-level waste to glass

International Nuclear Information System (INIS)

Beary, M.M.; Chick, L.A.; Ely, P.C.; Gott, S.A.

1977-06-01

The flowsheets describe a process for converting waste removed from the Hanford underground waste tanks to more immobile form. The process involves a chemical separation of the radionuclides from industrial chemicals, and then making glass from the resulting small volume of highly radioactive waste. Removal of Sr, actinides, cesium, and technetium is discussed

Actinide partitioning-transmutation program final report. I. Overall assessment

International Nuclear Information System (INIS)

Croff, A.G.; Blomeke, J.O.; Finney, B.C.

1980-06-01

This report is concerned with an overall assessment of the feasibility of and incentives for partitioning (recovering) long-lived nuclides from fuel reprocessing and fuel refabrication plant radioactive wastes and transmuting them to shorter-lived or stable nuclides by neutron irradiation. The principal class of nuclides considered is the actinides, although a brief analysis is given of the partitioning and transmutation (P-T) of 99 Tc and 129 I. The results obtained in this program permit us to make a comparison of the impacts of waste management with and without actinide recovery and transmutation. Three major conclusions concerning technical feasibility can be drawn from the assessment: (1) actinide P-T is feasible, subject to the acceptability of fuels containing recycle actinides; (2) technetium P-T is feasible if satisfactory partitioning processes can be developed and satisfactory fuels identified (no studies have been made in this area); and (3) iodine P-T is marginally feasible at best because of the low transmutation rates, the high volatility, and the corrosiveness of iodine and iodine compounds. It was concluded on the basis of a very conservative repository risk analysis that there are no safety or cost incentives for actinide P-T. In fact, if nonradiological risks are included, the short-term risks of P-T exceed the long-term benefits integrated over a period of 1 million years. Incentives for technetium and iodine P-T exist only if extremely conservative long-term risk analyses are used. Further RD and D in support of P-T is not warranted

Final waste classification and waste form technical position papers

International Nuclear Information System (INIS)

1983-05-01

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

Research needs in cement-based waste forms

International Nuclear Information System (INIS)

McDaniel, E.W.; Spence, R.D.; Tallent, O.K.

1990-01-01

Cement-based waste forms are one of the most widely used waste disposal options, yet definitive knowledge of the fate of the waste species inside the waste form is lacking. A fundamental understanding of the chemistry and microstructure of the waste forms would lead to a better understanding of the mass transfer of the waste species, more confidence in predicting and extrapolating waste form performance, and design of better waste forms. Better and cheaper leach tests would lead to quicker and more cost effective screening of waste form alternatives. In addition, assessment of durability may be important to predicting waste form performance in the field. It should be noted that the research needs discussed in this report are from the perspective of investigators working in applied waste management areas, while the proposed investigations are fundamental or basic. Details as to experimental methods and tools to be used in achieving the objectives of the proposed are research beyond the scope of this paper and are better filled in by others. In broad terms, the research topics discussed are correlation of cement-based waste form physical properties to performance, waste-form fundamental chemistry and microstructure, and product performance testing

Comparative assessment of TRU waste forms and processes. Volume I. Waste form and process evaluations

International Nuclear Information System (INIS)

Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.

1982-09-01

This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 10 5 per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables

Ceramic and glass radioactive waste forms

Energy Technology Data Exchange (ETDEWEB)

Readey, D.W.; Cooley, C.R. (comps.)

1977-01-01

This report contains 14 individual presentations and 6 group reports on the subject of glass and polycrystalline ceramic radioactive waste forms. It was the general consensus that the information available on glass as a waste form provided a good basis for planning on the use of glass as an initial waste form, that crystalline ceramic forms could also be good waste forms if much more development work were completed, and that prediction of the chemical and physical stability of the waste form far into the future would be much improved if the basic synergistic effects of low temperature, radiation and long times were better understood. Continuing development of the polycrystalline ceramic forms was recommended. It was concluded that the leach rate of radioactive species from the waste form is an important criterion for evaluating its suitability, particularly for the time period before solidified waste is permanently placed in the geologic isolation of a Federal repository. Separate abstracts were prepared for 12 of the individual papers; the remaining two were previously abstracted.

PREPARATION OF ACTINIDE-ALUMINUM ALLOYS

Science.gov (United States)

Moore, R.H.

1962-09-01

BS>A process is given for preparing alloys of aluminum with plutonium, uranium, and/or thorium by chlorinating actinide oxide dissolved in molten alkali metal chloride with hydrochloric acid, chlorine, and/or phosgene, adding aluminum metal, and passing air and/or water vapor through the mass. Actinide metal is formed and alloyed with the aluminum. After cooling to solidification, the alloy is separated from the salt. (AEC)

Neutronics design study on a minor actinide burner for transmuting spent fuel

International Nuclear Information System (INIS)

Choi, Hang Bok

1998-08-01

A liquid metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors. The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the doppler coefficient, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200 MWth core is able to transmute the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics. (author). 34 refs., 22 tabs., 14 figs

Conditioning of radioactive waste solutions by cementation

International Nuclear Information System (INIS)

Vejmelka, P.; Rudolph, G.; Kluger, W.; Koester, R.

1992-02-01

For the cementation of the low and intermediate level evaporator concentrates resulting from the reprocessing of spent fuel numerous experiments were performed to optimize the waste form composition and to characterize the final waste form. Concerning the cementation process, properties of the waste/cement suspension were investigated. These investigations include the dependence of viscosity, bleeding, setting time and hydration heat from the waste cement slurry composition. For the characterization of the waste forms, the mechanical, thermal and chemical stability were determined. For special cases detailed investigations were performed to determine the activity release from waste packages under defined mechanical and thermal stresses. The investigations of the interaction of the waste forms with aqueous solutions include the determination of the Cs/Sr release, the corrosion resistance and the release of actinides. The Cs/Sr release was determined in dependence of the cement type, additives, setting time and sample size. (orig./DG) [de

Removal of actinide elements from liquid scintillation cocktail wastes using liquid-liquid extraction and demulsification techniques

International Nuclear Information System (INIS)

Foltz, K.; Landsberger, S.; Srinivasan, B.; Vandegrift, G.F.

1994-01-01

For many years liquid scintillation cocktail (LSC) wastes have been generated and stored at Argonne National Laboratory (ANL). These wastes are stored in thousands of 10--20 m scintillation vials, many of which contain elements with Z > 88. Because storage space is limited, disposal of this waste is pressing. These wastes could be commercially incinerated if the radionuclides with Z>88 are reduced to sufficiently low levels. However, there is currently no deminimus level for these radionuclides, and separation techniques are still being tested. The University of Illinois is conducting experiments to separate radionuclides with Z > 88 from simulated LSC wastes by using liquid-liquid extraction (LLX) and demulsification techniques. The actinide elements are removed from the LSC by extraction into an aqueous phase after the cocktail has been demulsified. The aqueous and organic phases are separated and the organic phase, now free from radionuclides with Z > 88, can be sent to a commercial incineration facility. The aqueous phase may be treated and disposed of using existing techniques. The LLX separation techniques used solutions of sodium oxalate, aluminum nitrate, and tetrasodium EDTA at varying concentrations. These extractants were mixed with the simulated waste in a 1:1 volume ratio. Using 1.0M Na 4 EDTA salt solutions, decontamination ratios as high as 230 were achieved

LLNL SFA OBER SBR FY17 Program Management and Performance Report: Subsurface Biogeochemistry of Actinides

Energy Technology Data Exchange (ETDEWEB)

Kersting, Annie B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-06-23

A major scientific challenge in environmental sciences is to identify the dominant processes controlling actinide transport in the environment. It is estimated that currently, over 2200 metric tons of anthropogenic plutonium (Pu) has accumulated worldwide, a number that increases yearly with additional spent nuclear fuel (Ewing et al., 2010). Plutonium has been shown to migrate on the scale of kilometers, giving way to a critical concern that the fundamental biogeochemical processes that control its behavior in the subsurface are not well understood (Kersting et al. 1999; Novikov et al. 2006; Santschi et al. 2002). Neptunium (Np) is less prevalent in the environment; however, it is predicted to be a significant long-term dose contributor in high-level nuclear waste. Our focus on Np chemistry in this Science Plan is intended to help formulate a better understanding of Pu redox transformations in the environment and clarify the differences between the two long-lived actinides. The research approach of our Science Plan combines (1) Fundamental Mechanistic Studies that identify and quantify biogeochemical processes that control actinide behavior in solution and on solids, (2) Field Integration Studies that investigate the transport characteristics of Pu and test our conceptual understanding of actinide transport, and (3) Actinide Research Capabilities that allow us to achieve the objectives of this Scientific Focus Area (SFA) and provide new opportunities for advancing actinide environmental chemistry. These three Research Thrusts form the basis of our SFA Science Program.

Plutonium Management, Minor Actinides Partitioning and Transmutation R and D in France

International Nuclear Information System (INIS)

Cavedon, Jean-Marc; Courtois, Charles

2003-01-01

Jean-Marc Cavedon (CEA, France) then presented the developments concerning Plutonium management and minor actinides P and T research and development in France. By the 1991 law on high-level long-lived radioactive waste a research programme was launched in the areas: (i) geological disposal, (ii) conditioning and long-term storage, and (iii) radiotoxicity reduction by P and T. The results of the work in these areas will be presented to the French Government and Parliament in 2006. The control of Plutonium stocks generated by the French PWRs is proposed to increase Plutonium consumption in reactors and minimise radioactive waste production, and requires the recycling of actinides, especially Plutonium. In the long term, CEA intends to develop a new technology based on gas cooled reactors and their associated fuel cycle, including multiple recycling of Plutonium. The advantages of this development consist in the optimisation of the use of natural resources and the concentration of Plutonium in limited quantities of fuel rods. If needed, the minor actinides could also be recycled. The planned CEA developments depend on new fuel types and will lead to novel waste types (light glasses) with a reduction of long-term radiotoxicity. Radiotoxicity reductions by a factor of 3 to 5 are expected for Plutonium recycling scenarios, and by up to a factor of a few hundreds for Plutonium and minor actinides recycling scenarios. This gain is nearly independent on the reactor type used, but needs about 100 years of application to become effective in terms of making a difference in the total waste inventory to be disposed of

Synthesis of selective extractor for minor actinide elements

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Seung [Konyang University, Nonsan (Korea); Cho, Moon Hwan [Kangwon National University, Chunchon (Korea)

1998-04-01

To selectively co-separate the lanthanide and actinide elements (MA) such as Am or Cm ion from radioactive waste, synthesis of diamide derivatives has been accomplished. In addition, picoline amide derivatives were also synthesized for selectively separate the minor actinide elements from lanthanide elements. The content of research has don are as follows: (1) synthesis of diamide as co-extractant (2) introduction of n-tetradecyl to increase the lipophilicity (3) Picolyl chloride, intermediate of the final product, was synthesized by improved method rather than reported method. (4) The length of alkyl side chain was adjusted to increase the lipophilicity of free ligand and its derivatives able to selectively separate the actinide metal from lanthanide metal ions was successfully synthesized and determined their purity by analytical instruments. (author). 12 refs., 28 figs.

Nuclear fuel activity with minor actinides after their useful life in a BWR

International Nuclear Information System (INIS)

Martinez C, E.; Ramirez S, J. R.; Alonso V, G.

2016-09-01

Nuclear fuel used in nuclear power reactors has a life cycle, in which it provides energy, at the end of this cycle is withdrawn from the reactor core. This used fuel is known as spent nuclear fuel, a strong problem with this fuel is that when the fuel was irradiated in a nuclear reactor it leaves with an activity of approximately 1.229 x 10 15 Bq. The aim of the transmutation of actinides from spent nuclear fuel is to reduce the activity of high level waste that must be stored in geological repositories and the lifetime of high level waste; these two achievements would reduce the number of necessary repositories, as well as the duration of storage. The present work is aimed at evaluating the activity of a nuclear fuel in which radioactive actinides could be recycled to remove most of the radioactive material, first establishing a reference of actinides production in the standard nuclear fuel of uranium at end of its burning in a BWR, and a fuel rod design containing 6% of actinides in an uranium matrix from the enrichment tails is proposed, then 4 standard uranium fuel rods are replaced by 4 actinide bars to evaluate the production and transmutation of the same, finally the reduction of actinide activity in the fuel is evaluated. (Author)

Helium and fission gas behaviour in magnesium aluminate spinel and zirconia for actinide transmutation

NARCIS (Netherlands)

Damen, P.M.G.

2003-01-01

In order to reduce the long-term radiotoxicity of spent nuclear fuel, many studies are performed on partitioning and transmutation of actinides. In such a scenario, the long-lived radio-isotopes (mostly actinides) are partitioned from the nuclear waste, and subsequently transmuted or fissioned in a

EUROPART: an European integrated project on actinide partitioning

International Nuclear Information System (INIS)

Madic, C.; Baron, P.; Hudson, M.J.

2006-01-01

Full text of publication follows: The EUROPART project is a scientific integrated project between 24 European partners, from 10 countries, mostly funded by the European Community within the FP6, together with CRIEPI from Japan and ANSTO from Australia. EUROPART aims at developing chemical partitioning processes for the so-called minor actinides (MA) contained in nuclear wastes, i.e. from Am to Cf. In the case of the treatment of dedicated spent fuels or targets, the actinides to be separated also include U, Pu and Np. The techniques considered for the separation of these radionuclides belong to the fields of hydrometallurgy and pyrometallurgy, as in the previous European FP5 programs named PARTNEW, CALIXPART and PYROREP, respectively. The two main axes of research within EUROPART are: 1/ the partitioning of MA (from Am to Cf) from wastes issuing from the reprocessing of high burn-up UOX fuels and multi-recycled MOX fuels, 2/ the partitioning of the whole actinide family of elements for recycling, as an option for advanced dedicated fuel cycles (this work will be connected to the studies to be performed within the EUROTRANS European integrated project). In hydrometallurgy, the research is organized in five Work Packages (WP). Four are dedicated to the study of partitioning methods mainly based on the use of solvent extraction methods and of solid extractants, one WP is dedicated to the development of actinide co-conversion methods for fuel or target preparations. The research in pyrometallurgy is organized into four WPs, listed hereafter: (i) study of the basic chemistry of transuranium elements and of some fission products in molten salts (chlorides, fluorides), (ii) development of actinide partitioning methods, (iii) study of the conditioning of the salt wastes, (iv) system studies. Moreover, a strong management team is concerned not only with the technical and financial issues arising from EUROPART, but also with information, communication and benefits for Europe

Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

International Nuclear Information System (INIS)

Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

1976-05-01

For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

Standardized waste form test methods

International Nuclear Information System (INIS)

Slate, S.C.

1984-01-01

The Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results are presented: the radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach test methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-1 approved reference material are presented. Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transporation, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance

Standardized waste form test methods

International Nuclear Information System (INIS)

Slate, S.C.

1984-11-01

The Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results is presented: The radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach tests methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-1 approved reference material are presented. Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transportation, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance. 2 references, 2 figures

Combined Extraction of Cesium, Strontium, and Actinides from Alkaline Media: An Extension of the Caustic-Side Solvent Extraction (CSSX) Process Technology

International Nuclear Information System (INIS)

Kenneth Raymond

2004-01-01

The wastes present at DOE long-term storage sites are usually highly alkaline, and because of this, much of the actinides in these wastes are in the sludge phase. Enough actinide materials still remain in the supernatant liquid that they require separation followed by long-term storage in a geological repository. The removal of these metals from the liquid waste stream would permit their disposal as low-level waste and dramatically reduce the volume of high-level wastes

Chemistry research for the Canadian nuclear fuel waste management program

International Nuclear Information System (INIS)

Vikis, A.C.; Garisto, F.; Lemire, R.J.; Paquette, J.; Sagert, N.H.; Saluja, P.P.S.; Sunder, S.; Taylor, P.

1988-01-01

This publication reviews chemical research in support of the Canadian Nuclear Fuel Waste Management Program. The overall objective of this research is to develop the fundamental understanding required to demonstrate the suitability of waste immobilization media and processes, and to develop the chemical information required to predict the long-term behaviour of radionuclides in the geosphere after the waste form and the various engineered barriers containing it have failed. Key studies towards the above objective include experimental and theoretical studies of uranium dioxide oxidation/dissolution; compilation of thermodynamic databases and an experimental program to determine unavailable thermodynamic data; studies of hydrothermal alteration of minerals and radionuclide interactions with such minerals; and a study examining actinide colloid formation, as well as sorption of actinides on groundwater colloids

Incorporation of transuranic elements in titanate nuclear waste ceramics

International Nuclear Information System (INIS)

Matzke, H.J.; Ray, I.L.F.; Theile, H.; Trisoglio, C.; Walker, C.T.; White, T.J.

1990-01-01

The incorporation of actinide elements and their rare-earth element analogues in titanate nuclear waste forms in reviewed. New partitioning data are presented for three waste forms containing Purex waste simulant in combination with either NpO 2 , PuO 2 , or Am 2 O 3 . The greater proportion of transuranics partition between perovskite and zirconolite, while some americium may enter loveringite. Autoradiography revealed clusters of plutonium atoms which have been interpreted as unreacted dioxide or sesquioxide. It is concluded that the solid-state behavior of transuranic elements in titanate waste forms is poorly understood, certainly not well enough to tailor a ceramic for the incorporation of waste from reprocessing of fast breeder reactor fuel in which transuranic species are more abundant than in Purex waste

Partitioning of Mercury from Actinides in the TRUEX Process

International Nuclear Information System (INIS)

Fiskum, Sandra K.; Rapko, Brian M.; Lumetta, Gregg J.

2000-01-01

A mercury complexant, L-cysteine hydrochloride, was tested for use in separating Hg from actinides during TRUEX processing of wastes at the Idaho National Engineering and Environmental Laboratory (INEEL). Mercury, americium, plutonium and uranyl distributions from TRUEX solvent were characterized over a nitric acid concentration range of 0.01 to 2 M. The applicability of cysteine was also evaluated for selective Hg complexation in an INEEL sodium-bearing waste simulant. A test was also conducted to evaluate the applicability of cysteine to separate Hg from Sr in the SREX process with Sr Resin used as a stand-in for the SREX process solvent. In all cases, the use of L-cysteine HCl retained Hg in the aqueous phase while causing no or little perturbation in the actinide and Sr distribution behavior

Permanganate Treatment of Savannah River Site Simulant Wastes for Strontium and Actinide Removal

International Nuclear Information System (INIS)

Wilmarth, W.R.

2003-01-01

This study examined the use of sodium permanganate and strontium nitrate to remove the actinides and radio-strontium from Savannah River Site (SRS) waste supernate. We examined the quantities of chemical feed reagents along with increased mixing and the excess of organic reductant. Additionally, we examined two processing schemes including that applicable to either the Salt Waste Processing Facility or the Alpha Removal Process (ARP) (5.6 M sodium ion concentration) conditions and the conditions for an In-Tank application (7.5 M sodium ion concentration). Our results support the following conclusions: The process met minimum required decontamination factors (DFs) within the tested parameter sets for strontium and plutonium in both the ARP and In-Tank application. The strontium DFs far exceeded the required values within the tested parameter sets. Within the ARP application, the use of peroxide as the reductant for permanganate produced higher plutonium DFs than the use of sodium formate. Reductant concentration and degree of mixing strongly influenced radionuclide decontamination. In the formate application under the ARP process, increasing the reductant concentration and mixing energy resulted in higher Sr and Pu decontamination

Loading Actinides in Multilayered Structures for Nuclear Waste Treatment: The First Case Study of Uranium Capture with Vanadium Carbide MXene.

Science.gov (United States)

Wang, Lin; Yuan, Liyong; Chen, Ke; Zhang, Yujuan; Deng, Qihuang; Du, Shiyu; Huang, Qing; Zheng, Lirong; Zhang, Jing; Chai, Zhifang; Barsoum, Michel W; Wang, Xiangke; Shi, Weiqun

2016-06-29

Efficient nuclear waste treatment and environmental management are important hurdles that need to be overcome if nuclear energy is to become more widely used. Herein, we demonstrate the first case of using two-dimensional (2D) multilayered V2CTx nanosheets prepared by HF etching of V2AlC to remove actinides from aqueous solutions. The V2CTx material is found to be a highly efficient uranium (U(VI)) sorbent, evidenced by a high uptake capacity of 174 mg g(-1), fast sorption kinetics, and desirable selectivity. Fitting of the sorption isotherm indicated that the sorption followed a heterogeneous adsorption model, most probably due to the presence of heterogeneous adsorption sites. Density functional theory calculations, in combination with X-ray absorption fine structure characterizations, suggest that the uranyl ions prefer to coordinate with hydroxyl groups bonded to the V-sites of the nanosheets via forming bidentate inner-sphere complexes.

J-ACTINET activities of training and education for actinide science research

International Nuclear Information System (INIS)

Miato, Kazuo; Konashi, Kenji; Yamana, Hajimu; Yamanaka, Shinsuke; Nagasaki, Shinya; Ikeda, Yasuhisa; Sato, Seichi; Arita, Yuji; Idemitsu, Kazuya; Koyama, Tadafumi

2011-01-01

Actinide science research is indispensable to maintain sustainable development of innovative nuclear technology, especially advanced fuels, partitioning/reprocessing, and waste management. For actinide science research, special facilities with containment and radiation shields are needed to handle actinide materials since actinide elements are γ-, α- and neutron-emitters. The number of facilities for actinide science research has been decreased, especially in universities, due to the high maintenance cost. J-ACTINET was established in 2008 to promote and facilitate actinide science research in close cooperation with the facilities and to foster many of young scientists and engineers to be actively engaged in the fields of actinide science. The research program was carried out, through which young researchers were expected to learn how to make experiments with advanced experimental tools and to broaden their horizons. The summer schools and computational science school were held to provide students, graduate students, and young researchers with the opportunities to come into contact with actinide science research. In these schools, not only the lectures, but also the practical exercises were made as essential part. The overseas dispatch program was also carried out, where graduate students and young researchers were sent to the international summer schools and conferences. (author)

Combined Waste Form Cost Trade Study

International Nuclear Information System (INIS)

Gombert, Dirk; Piet, Steve; Trickel, Timothy; Carter, Joe; Vienna, John; Ebert, Bill; Matthern, Gretchen

2008-01-01

A new generation of aqueous nuclear fuel reprocessing, now in development under the auspices of the DOE Office of Nuclear Energy (NE), separates fuel into several fractions, thereby partitioning the wastes into groups of common chemistry. This technology advance enables development of waste management strategies that were not conceivable with simple PUREX reprocessing. Conventional wisdom suggests minimizing high level waste (HLW) volume is desirable, but logical extrapolation of this concept suggests that at some point the cost of reducing volume further will reach a point of diminishing return and may cease to be cost-effective. This report summarizes an evaluation considering three groupings of wastes in terms of cost-benefit for the reprocessing system. Internationally, the typical waste form for HLW from the PUREX process is borosilicate glass containing waste elements as oxides. Unfortunately several fission products (primarily Mo and the noble metals Ru, Rh, Pd) have limited solubility in glass, yielding relatively low waste loading, producing more glass, and greater disposal costs. Advanced separations allow matching the waste form to waste stream chemistry, allowing the disposal system to achieve more optimum waste loading with improved performance. Metals can be segregated from oxides and each can be stabilized in forms to minimize the HLW volume for repository disposal. Thus, a more efficient waste management system making the most effective use of advanced waste forms and disposal design for each waste is enabled by advanced separations and how the waste streams are combined. This trade-study was designed to juxtapose a combined waste form baseline waste treatment scheme with two options and to evaluate the cost-benefit using available data from the conceptual design studies supported by DOE-NE

IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS

International Nuclear Information System (INIS)

Bernot, P.

2004-01-01

The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The models are based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. The models described in this report constitute the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS:167796]) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2003 [DIRS: 166296]). The technical work plan was prepared in accordance with AP-2.27Q, Planning for Science Activities. Any deviations from the technical work plan are documented in the following sections as they occur. The TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model assessments: (1) Mechanical and thermal impacts of basalt magma intrusion on the invert, waste packages and waste forms of the intersected emplacement drifts of Zone 1. (2) Temperature and pressure trends of basaltic magma intrusion intersecting Zone 1 and their potential effects on waste packages and waste forms in Zone 2 emplacement drifts. (3) Deleterious volatile gases, exsolving from the intruded basalt magma and their potential effects on waste packages of Zone 2 emplacement drifts. (4) Post-intrusive physical

Hot test of a TALSPEAK procedure for separation of actinides and lanthanides using recirculating DTPA-lactic acid solution

International Nuclear Information System (INIS)

Persson, G.; Svantesson, I.; Wingefors, S.; Liljenzin, J.O.

1984-01-01

Results are reported from a hot test of a TALSPEAK type process for separation of higher actinides (Am, Cm) from lanthanides. Actinides and lanthanides are extracted by 1 M HDEHP and separated by selective strip of the actinides, using a mixture of DTPA and lactic acid (reversed TALSPEAK process). In order to minimize the generation of secondary waste, a procedure using recirculating DTPA-Lactic acid solution has been developed. A separation factor between Am and Eu of 132 was achieved. In regard to separations of Am and Cm from commercial HLLW (high level liquid wastes), the factor corresponds to 1.5% of the lanthanide group remaining with the actinides. The loss of Am was about 0.2%. 9 figures, 3 tables

Liquid Secondary Waste Grout Formulation and Waste Form Qualification

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-23

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

Mercury extraction by the TRUEX process solvent. II. Selective partitioning of mercury from co-extracted actinides in a simulated acidic ICPP waste stream

International Nuclear Information System (INIS)

Brewer, K.N.; Herbst, R.S.; Tranter, T.J.; Todd, T.A.

1995-01-01

The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/l. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, 203 Hg, was spiked into a simulated solution of SBW containing 1 g/l mercury. Successive extraction batch contacts with the mercury spiked waste and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl 2 , from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HNO 3 and 0.077 with 0.25 M Na 2 CO 3 . Because experiments described here show that mercury can be extracted from SBW and stripped from the solvent, a process has been developed to partition mercury from the actinides in SBW. 10 refs., 3 figs., 10 tabs

The INE-Beamline for Actinide Research at ANKA

Science.gov (United States)

Brendebach, Boris; Denecke, Melissa A.; Rothe, Jörg; Dardenne, Kathy; Römer, Jürgen

2007-02-01

The INE-Beamline for actinide research at the synchrotron source ANKA is now fully operational. This beamline was designed, built, and commissioned by the Institut für Nukleare Entsorgung (INE) at the Forschungszentrum Karlsruhe (FZK), Germany. It is dedicated to actinide speciation investigations related to nuclear waste disposal as well as applied and basic actinide research. Experiments on radioactive samples with activities up to 106 times the limit of exemption inside a safe and flexible double containment concept are possible. The close proximity of the beamline to INE's active laboratories is unique in Europe. Currently, experiments can be performed in an X-ray energy range from around 2.15 keV (P K edge) to 24.35 keV (Pd K edge). The INE-Beamline design is optimized for spectroscopy with emphasis on surface sensitive techniques. A microfocus option is presently being installed at the INE-Beamline. Access to the INE-Beamline is possible through cooperation with INE, through the ANKA proposal system and via the European Network of Excellence for Actinide Sciences (ACTINET).

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

Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Special waste-form lysimeters: Arid

International Nuclear Information System (INIS)

Jones, T.L.; Serne, R.J.

1987-08-01

The release of contaminant from solidified low-level waste forms is being studied in a field lysimeter facility at the Hanford Site in southeastern Washington State. Duplicate samples of five different waste forms have been buried in 10 lysimeters since March 1984. Waste-form samples represent three different waste streams and four solidification agents (masonry cement, Portland III cement, Dow polymer /sup (a)/, and bitumen). Most precipitation at the Hanford Site arrives as winter snow; this contributes to a strong seasonal pattern in water storage and drainage observed in the lysimeters. The result is an annual range in the volumetric soil water content from 11% in late winter to 7% in the late summer and early fall, as well as annual changes in pore water velocities from approximately 1 cm/wk in early spring to less than 0.05 cm/wk in early fall. Measurable quantities of tritium and cobalt-60 are being collected in lysimeter drainage water. Approximately 30% of the original tritium inventory has been leached from two lysimeters originally containing tritium. Cobalt-60 is present in all waste forms; it is being collected in the leachate from five lysimeters. The total amount released varies, but in each case it is less than 0.1% of the original cobalt inventory of the waste sample. Nonradioactive constituents contained in the waste form, such as sodium, boron, and sulfate, are also being leached

Actinide recycling in reactors; Aktiniden-Rezyklierung in Reaktoren

Energy Technology Data Exchange (ETDEWEB)

Kuesters, H.; Wiese, H.W.; Krieg, B.

1995-08-01

The objective is an assessment of the transmutation of long-lived actinides and fission products and the incineration of plutonium for reducing the risk potential of radioactive waste from reactors in comparison to direct waste disposal. The contribution gives an interim account on homogeneous and heterogeneous recycling of `risk nuclides` in thermal and fast reactors. Important results: - A homogeneous 5 percent admixture of minor actinides (MA) from N4-PWRs to EFR fuel would allow a transmutation not only of the EFR MA, but in addition of the MA from 5 or 6 PWRs of equal power. However, the incineration is restricted by safety considerations. - LWR have only a very low MA incineration potential, due to their disadvantageous neutron capture/fission ratio. - In order to keep the Cm inventory at a low level, it is advantageous to concentrate the Am heterogeneously in particular fuel elements or rods. (orig./HP)

Extraction of actinides from chloride medium using pentaalkylpropanediamides

International Nuclear Information System (INIS)

Cuillerdier, C.; Musikas, C.

1991-01-01

Pyrometallurgical processes for the purification of plutonium create waste solutions containing actinides, mainly americium, in chloride medium. Studies have been undertaken to study the extraction of actinides in chloride medium (hydrochloric acid mixed with concentrated salts such as LiCl, CaCl 2 , MgCl 2 , KCl) using pentaalkylpropanediamides as extractants. Plutonium (IV) is very easily extracted, Am (III) needs a salting out agent such as LiCl. Back extraction of trivalent cations is easy in HCl <5M. Plutonium(IV) and (VI) can be stripped by reduction either with ascorbic acid or hydroxylammonium salts in weak acid medium. Several diluents can be used (aromatic, chlorinated or even aliphatic) with addition of decanol to prevent third phase formation. In conclusion diamides can be used for various wastes declassification, they are potentially completely incinerable, and, as the synthesis has been optimized, they appear to be promising extractants

Recovery of the transplutonium elements from nuclear reactor waste

International Nuclear Information System (INIS)

Campbell, D.O.; Buxton, S.R.

1980-01-01

A new method for separating actinides from fission product waste does not require quantitative precipitation of radioactive solids. The method involves a refinement of the first partitioning step in a two-step process. The nitric acid content of the reprocessing waste stream, which contains actinides, lanthanides and other metals, is first adjusted to 0.1 - 1.0 M. An excess of oxalate ions is added to cause most of the actinides and lanthanides to precipitate as solid oxalates leaving a supernate solution containing the remaining actinides and lanthanides and other metals. The oxalate precipitate is separated from the precipitation zone to provide a trivalent fraction for the second actinide/lanthanide partitioning step. The supernatant solution is contacted with sufficiently strong acid ion exchange resin so that the actinides and lanthanides are adsorbed. The raffinate is substantially free of actinides. The actinides and lanthanides are eluted from the resin with 3-6 M nitric acid. The eluate may be concentrated and combined with the trivalent fraction obtained prior to the actinide/lathanide partitioning step. Or, the actinides and lanthanides from the eluate can be combined with additional reprocessing waste for recycle. (DN)

A U-bearing composite waste form for electrochemical processing wastes

Energy Technology Data Exchange (ETDEWEB)

Chen, X.; Ebert, W. L.; Indacochea, J. E.

2018-04-01

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.

Hot and cold pressing of (La,Ce)PO4-based nuclear waste forms

International Nuclear Information System (INIS)

Floran, R.J.; Rappaz, M.; Abraham, M.M.; Boatner, L.A.

1981-01-01

Synthetic analogs of the mineral monazite [(Ce,La,Ca,Th,U)(P,Si)O 4 ] are promising host phases for the isolation of actinide wastes. In the present investigation, the optimal conditions necessary to form high-density, simulated waste pellets from calcined LaPO 4 and CePO 4 powders have been examined. Pellets that are close to the theoretical density are necessary in order to minimize porosity and hence potential avenues along which fluid/waste interactions can take place. Calcined powders prepared by a urea precipitation process were cold pressed and hot pressed under a range of controlled conditions to form coherent, cylindrical pellets. Changes in density were examined as a function of pressure, temperature and duration of sintering. For cold-pressed pellets, a significant increase in density occurs during sintering between 1000 0 C and 1100 0 C. This increase is correlated with substantial grain coarsening of the pellet microstructure. A comparison of cold and hot pressing techniques suggests that, only after sintering, does the density of cold-pressed pellets approach (but not equal) that of the hot-pressed pellets. Densities >90% of the theoretical value of 5.11 g/cm 3 are easily attainable by hot pressing without sintering. The significance of density differences on pellet stability will be investigated in future leaching studies. The apparent advantages of the higher densities achieved by hot pressing must be weighed against the increased technological/engineering complexities involved when working at sustained high temperatures in a remote environment. Thus the cold pressing technique may ultimately prove more practical for large-scale commercial operations

Survey of concrete waste forms

International Nuclear Information System (INIS)

Moore, J.G.

1981-01-01

The incorporation of radioactive waste in cement has been widely studied for many years. It has been routinely used at nuclear research and production sites for some types of nuclear waste for almost three decades and at power reactor plants for nearly two decades. Cement has many favorable characteristics that have contributed to its popularity. It is a readily available material and has not required complex and/or expensive equipment to solidify radioactive waste. The resulting solid products are noncombustible, strong, radiation resistant, and have reasonable chemical and thermal stability. As knowledge increased on the possible dangers from radioactive waste, requirements for waste fixation became more stringent. A brief survey of some of the research efforts used to extend and improve cementitious waste hosts to meet these requirements is given in this paper. Selected data are presented from the rather extensive study of the applicability of concrete as a waste form for Savannah River defense waste and the use of polymer impregnation to reduce the leachability and improve the durability of such waste forms. Hot-pressed concretes that were developed as prospective host solids for high-level wastes are described. Highlights are given from two decades of research on cementitious waste forms at Oak Ridge National Laboratory. The development of the hydrofracture process for the disposal of all locally generated radioactive waste led to a process for the disposal of I-129 and to the current research on the German in-situ solidification process for medium-level waste and the Oak Ridge FUETAP process for all classes of waste including commercial and defense high-level wastes. Finally, some of the more recent ORNL concepts are presented for the use of cement in the disposal of inorganic and biological sludges, waste inorganic salts, trash, and krypton

Interface reactions of actinides on muscovite; Grenzflaechenreaktionen von Actiniden an Muskovit

Energy Technology Data Exchange (ETDEWEB)

Hellebrandt, Stefan

2017-11-06

This thesis investigates the interaction of actinides with siliceous minerals on the molecular level to contribute to a reliable geochemical long-term safety assessment for a nuclear waste disposal site. The isolation in deep geological formations is the globally preferred strategy for the disposal of actinides like plutonium or americium. The reason is that their high radiotoxicity, due to their high radioactivity and long half-life, requires isolation from the biosphere over a time period up to 1 million years. It is necessary to study the behavior of radionuclides in such a waste disposal site and its surroundings to predict their mobility in the case of a release due to water ingress, and to guarantee that aquifers and the biosphere are not affected. The decisive parameter is the mobility of the radionuclides, which itself depends on different retention mechanisms. One of these is the sorption on mineral phases. The multi barrier system of a waste disposal site provides a lot of different mineral phases: In the geotechnical barrier are for example swellable clay minerals as bentonite, the geological barrier consists of different mineral phases depending on the host rock (granite, claystone or salt). To improve the understanding of sorption processes, their mechanisms, and what reaction parameters influence them, is the content of the present work. To obtain this understanding on a molecular level, modern analytical techniques must be applied. Mainly, Surface X-ray Scattering (SXS), with Crystal Truncation Rod (CTR) and Resonant Anomalous X-ray Reflectivity (RAXR) measurements, and site-selective time-resolved laser fluorescence spectroscopy (TRLFS), were combined with alpha-spectrometry and grazing incidence X-ray absorption near-edge structure (GI-XANES) spectroscopy to obtain and validate results in great detail. Compared to previous findings, these results are consistent with the adsorption of Pu in the form of Pu(IV)-oxo-nanoparticles. In contrast, no

Mass spectrometric studies of the complexing behaviour of actinide ions in solution

International Nuclear Information System (INIS)

Steppert, Michael

2012-01-01

As the long-term radiotoxicity of spent nuclear fuel is governed by Plutonium and the Minor Actinides, these elements are focussed on for investigations in the framework of safety assessment for nuclear waste repositories. To shed more light on the selectivity of the partitioning ligands BTP and BTBP towards the extraction of trivalent actinides, the complexes these ligands form with lanthanides in octanolic solution were characterized. The differences in the extraction efficiencies among the different lanthanides were traced back to the varying preferential formation of Ln(BTP)3 complexes, depending on the ionic radius of the lanthanides. Additionally it was shown that depending on the sterical demand of BTBP ligands nitrate anions coordinate in the first coordination shell of Eu(BTBP)2-complexes. As the behavior of Plutonium under geochemical conditions is of particular interest for the safety assessment of potential nuclear waste repositories, the second part of the thesis focuses on the hydrolysis and colloid formation behavior of aqueous Plutonium solutions. The solution species of Zirconium(IV) as analogue for Plutonium(IV) as well as of Uranium(VI) and Plutonium(VI) were characterized and quantified by means of electrospray ionization mass spectrometry. Moreover the colloid-induced reduction of Pu(V) to Pu(IV) and the subsequent formation of colloidal species was investigated. [de

Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

Energy Technology Data Exchange (ETDEWEB)

Choi, Jongkwon [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Um, Wooyong, E-mail: wooyong.um@pnnl.gov [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Choung, Sungwook [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of)

2014-09-15

The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl–KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl–KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl–KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl–KCl eutectic salt and off-gas scrubber solutions with significant cost savings.