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Sample records for synroc

  1. Synroc processing options

    International Nuclear Information System (INIS)

    Rozsa, R.B.; Hoenig, C.L.

    1981-01-01

    Synroc is a titanate-based ceramic material currently being developed for immobilizing high-level nuclear reactor wastes in solid form. Synroc D is a unique variation of Synroc. It can contain the high-level defense wastes, particularly those in storage at the Savannah River Plant. In this report, we review the early development of the initial Synroc process, discuss modification and other options that simplify it overall, and recommend the future direction of research and development in the processing area. A reference Synroc process is described briefly and contrasted with the Savannah River Laboratory glass-based reference case. Preliminary engineering layouts show Synroc to be a more complex processing operation and, thus, more expensive than the glass-based process. However, we believe that simplifications, which will significantly reduce the cost difference, are possible. Further research and development will continue in the areas of slurry processing, fluidized bed calcination, and mineralization. This last will use sintering, hot uniaxial pressing, or hot isostatic pressing

  2. Sintering of Synroc D

    International Nuclear Information System (INIS)

    Robinson, G.

    1982-01-01

    Sintering has been investigated as a method for the mineralization and densification of high-level nuclear defense waste powder. Studies have been conducted on Synroc D composite powder LS04. Optimal densification has been found to be highly dependent on the characteristics of the starting material. Powder subjected to milling, which was believed to reduce the level of agglomeration and possibly particle size, was found to densify better than powder not subjected to this milling. Densities of greater than 95% of theoretical could be achieved for samples sintered at 1150 to 1200 0 C. Mineralogy was found to be as expected for Synroc D for samples sintered in a CO 2 /CO atmosphere where the Fe +2 /Fe +3 ratio was maintained at 1.0 to 5.75. In a more oxidizing, pure CO 2 atmosphere a new phase, not previously identified in Synroc D, was found

  3. Nuclear waste immobilisation in SYNROC

    International Nuclear Information System (INIS)

    Ringwood, A.E.

    1984-04-01

    SYNROC is a crystalline titanate ceramic designed to immobilise the elements occurring in high level wastes. It has been demonstrated that the great majority of elements present in high level wastes can be incorporated within the crystalline lattices of the SYNROC minerals. In this state they are extremely resistant to attack by aqueous solutions. Extensive experimental data demonstrates that SYNROC is 1,000 to 10,000 times more resistant to leaching than borosilicate glass wasteforms at 100 - 200 deg C. SYNROC displays exceptional stability at higher temperatures where glasses disintegrate rapidly. The essential minerals of SYNROC occur in nature where they have demonstrated their capacity to survive in a wide range of geological and geochemical environments for periods of 10 8 - 10 9 years. These characteristics, in combination with the experimental studies, demonstrate that SYNROC offers important advantages over borosilicate glass as a wasteform, both in terms of performance and capacity to achieve public acceptability. Studies of the properties of ancient naturally occurring SYNROC minerals containing uranium and thorium which have received very large cumulative radiation doses demonstrate that the capacity of these minerals to retain waste elements is not substantially retarded by radiation damage. Process technology for the production of SYNROC on a large scale is now under development. A novel method employing uniaxial hot pressing of SYNROC powder contained in free sanding steel bellows at 1150 deg C yields a fully dense product. Production costs are estimated to be in the same range as for borosilicate glass

  4. Synroc for plutonium disposal

    International Nuclear Information System (INIS)

    Johnston, A.; Vance, E.R.

    1999-01-01

    A pyrochlore-rich titanate ceramic has been chosen by the US DOE for excess weapons Pu immobilisation in the USA. The development of this wasteform was based on the Synroc strategy which aims to immobilise radioactive waste in durable multiphase titanate ceramics with phases chosen to he similar to titanate minerals that exist in nature and have immobilised U and Th for billions of years. The evolution of the pyrochlore-rich ceramic for Pu immobilisation from earlier Synroc variants is described and the choice of process steps is discussed. Leaching studies demonstrate that the release rate of Pu from the wasteforms in aqueous media is very low and similar to those of U and the neutron absorbers Gd and Hf that will ensure avoidance of nuclear criticality in repository environments

  5. Status of the synroc project

    International Nuclear Information System (INIS)

    Reeve, K.D.; Ramm, E.J.; Woolfrey, J.L.; Ryan, R.K.; Buykx, W.J.; Cassidy, D.J.; Webb, C.E.

    1980-10-01

    SYNROC-B has been proposed as a vehicle for the immobilisation of solidified radioactive waste. It consists of an assemblage of three synthetic mineral phases: perovskite, barium hollandite and zirconolite. Fabrication studies, leach testing and irradiation testing of SYNROC are reported

  6. Synroc - progress and future prospects

    International Nuclear Information System (INIS)

    Jostsons, A.

    2001-01-01

    Most of the early development of SYNROC focused on the SYNROC-C formulation for immobilising liquid HLW from the reprocessing of commercial LWR spent fuel. Subsequently, ANSTO has responded to developments in R and D on partitioning and transmutation, excess plutonium disposition and the needs of global remediation programs by developing a variety of titanate ceramic waste forms for specific applications. This paper reviews the progress in the development of titanate ceramics and ceramic/glass composites and addresses the relevance of this work in future radioactive waste management strategies

  7. Nuclear waste locked up in Synroc

    International Nuclear Information System (INIS)

    Grose, S.

    1998-01-01

    Australian technology Synroc leads the way in immobilizing nuclear waste. The ceramic approach set Synroc apart from glass-based technologies being developed in the US and Europe. Despite its international recognition, local industry failed to turn it into a commercial product. The author warns that if ANSTO is loosing the capacity to maintain Australian involvement in the development of Synroc, there is a danger that overseas interests would pick up the profit steam from this great Australian innovation

  8. Fluid-bed process for SYNROC production

    International Nuclear Information System (INIS)

    Ackerman, F.J.; Grens, J.Z.; Ryerson, F.J.; Hoenig, C.L.; Bazan, F.; Peters, P.E.; Smith, R.; Campbell, J.H.

    1983-01-01

    SYNROC is a titanate-based ceramic waste developed for the immobilization of high-level nuclear reactor waste. Lawrence Livermore National Laboratory (LLNL) has investigated a fluid-bed technique for the large-scale production of SYNROC precursor powders. Making SYNROC in a fluid bed permits slurry drying, calcination and reduction-oxidation reactions to be carried out in a single unit. We present the results of SYNROC fluid-bed studies from two fluid-bed units 10 cm in diameter: an internally heated fluid-bed unit developed by Exxon Idaho and an externally heated unit constructed at LLNL. Bed operation over a range of temperatures, feed rates, fluidizing rates, and redox conditions indicate that SYNROC powders of a high density and a uniform particle size can be produced. These powders facilitate the densification step and yield dense ceramics (greater than 95% theoretical density) with well-developed phases and low leaching rates

  9. AAEC builds synroc demonstration plant

    International Nuclear Information System (INIS)

    O'Hagan, R.

    1986-01-01

    A demonstration plant to test the feasibility of an Australian-developed method of immobilising radioactive waste is being built at the Australian Atomic Energy Commission's Lucas Heights Research Laboratories. The plant will operate as if radioactive waste was actually being processed, but non-radioactive elements of a similar composition will be used. The process involves the simulated waste being mixed into a slurry with the main SYNROC ingredients and then converted to a powder. The powder is moved about the plant in bellows-type containers by robots

  10. Solution chemistry techniques in SYNROC preparation

    International Nuclear Information System (INIS)

    Dosch, R.G.; Lynch, A.W.

    1981-07-01

    Investigations of titanate-based ceramic forms for radioactive waste immobilization are underway at Sandia National Laboratories (SNLA) and at Lawrence Livermore National Laboratory (LLNL). Although the waste forms differ as to overall product composition, the waste-containing phases in both ceramic products have similar crystalline structure types. These include metallic phases along with oxides with structure types of the mineral analogues perovskite, zirconolite, and hollandite. Significant differences also exist in the area of processing. More conventional ceramic processing methods are used at LLNL to produce SYNROC while solution chemistry techniques involving metal alkoxide chemistry and ion exchange have been developed at SNLA to prepare calcium titanate-based waste ceramics. The SNLA techniques were recently modified and applied to producing SYNROC (compositions C and D) as part of an interlaboratory information exchange between SNLA and LLNL. This report describes the methods used in preparing SYNROC including the solution interaction, and hot-pressing methods used to obtain fully dense SYNROC monoliths

  11. Microstructural aspects of SYNROC from sandia precursor

    International Nuclear Information System (INIS)

    Stevens, G.T.; Watson, K.G.; Bellrose, A.

    1987-04-01

    Typical microstructures formed in Synroc C have been observed by optical and scanning electron microscopy. The principal effects of segregation, variation of calciner atmosphere and change in hot-pressing conditions, are summarised

  12. An assessment of the Synroc process

    International Nuclear Information System (INIS)

    Harley, P.E.; Birch, D.

    1981-11-01

    A study has been carried out of proposed routes for the full scale manufacture of Synroc [incorporation of elements occurring in radioactive wastes into several mineral phases so as to produce a synthetic rock] with an appreciation of the technologies involved in the various stages of manufacture. Possible problem areas have been identified and solutions suggested. To provide a baseline for the Synroc assessment a comparison has been made with first generation borosilicate glass processes. A conceptual model Synroc plant has been specified and areas where further research and development are required to establish the feasibility of the process have been identified. Emphasis throughout the study has been placed on the plant rather than on the product although for the sake of completeness Appendix A to the report describes current knowledge of the properties of Synroc. (author)

  13. Calcination under negative atmosphere for SYNROC preparation

    International Nuclear Information System (INIS)

    Ambashta, R.D.; Wattal, P.K.; Govindankutty, K.V.

    2006-01-01

    SYNROC-C is a ceramic waste formulation designed to immobilise reprocessing waste from fast breeder reactor. This formulation is capable of incorporating noble metals, other fission products, corrosion products and activation products in its multiphase assemblage. Calcination is an important step of SYNROC preparation for decomposition of nitrates of the radioactive waste and conversion to oxide precursors. This paper presents a comparison between properties of calcine prepared under different calcination procedures to obtain product suitable for compaction

  14. Synroc tailored waste forms for actinide immobilization

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  15. SYNROC production using a fluid bed calciner

    International Nuclear Information System (INIS)

    Ackerman, F.J.; Grens, J.Z.; Ryerson, F.J.; Hoenig, C.L.; Bazan, F.; Campbell, J.H.

    1982-01-01

    SYNROC is a titanate-based ceramic developed for immobilization of high-level nuclear reactor wastes in solid form. Fluid-bed SYNROC production permits slurry drying, calcining and redox to be carried out in a single unit. We present results of studies from two fluid beds; the Idaho Exxon internally-heated unit and the externally-heated unit constructed at Lawrence Livermore National laboratory. Bed operation over a range of temperature, feed rate, fluidizing rate and redox conditions indicate that high density, uniform particle-size SYNROC powders are produced which facilitate the densification step and give HUP parts with dense, well-developed phases and good leaching characteristics. 3 figures, 3 tables

  16. SYNROC C: preparation and radwaste distribution

    International Nuclear Information System (INIS)

    Ryerson, F.J.; Hoenig, C.L.; Smith, G.S.

    1981-01-01

    Results of the synthesis of SYNROC C from both high surface area (16m 2 /g) and low surface area (4m 2 /g) powders at low and high oxygen fugacities indicate variations in radionuclide distribution. These results are most striking for the partitioning of uranium between perovskite and zirconolite. In highly active powders, the formation of a pyrochlore precursor ensures that equilibrium partitioning is approached. In less active powders, no low temperature pyrochlore is formed. As a result, the uranium partitioning is a function of nucleation processes. At high oxygen fugacities, uranium is rejected from perovskite and an additional uranium rich phase is formed precluding the synthesis of SYNROC C in air

  17. Microwave heating application in calcination and SYNROC formation

    International Nuclear Information System (INIS)

    Ambashta, R.D.; Wattal, P.K.; Malav, R.K.; Mallik, G.K.

    2006-01-01

    Microwave for calcination of titanate based ceramic wasteform (SYNROC) is being reported for the first time in this paper. Although major constituents in SYNROC were non microwave active, the combination with microwave active constituents rendered the mixture calcinable. Calcine was sintered at 1150 degC under hot uniaxial conditions at an applied pressure of ∼30 MPa. XRD shows presence of major phases of SYNROC in the compacted sample. (author)

  18. Preliminary studies on fabrication routes for SYNROC

    International Nuclear Information System (INIS)

    Evans, J.P.; Paige, E.L.

    1980-12-01

    The use of Synroc as a disposal medium for magnox waste has been evaluated and three possible methods of fabrication have been investigated. Hot pressing in graphite dies has produced the highest densities - further work is proceeding on sintering and hot isostatic pressing. The leach test results have indicated that the lowest density samples have adequate leach resistance while the higher density samples are an order of magnitude better. (author)

  19. Heat transfer in large compacts of SYNROC powder

    International Nuclear Information System (INIS)

    Buykx, W.J.

    1984-01-01

    The parameters determining the time required to reach temperature uniformity in a shock heated cylindrical compact of SYNROC powder are identified as the dimensions of the compact and the thermal diffusivity of the material. The effect of shape and size of the compact are discussed, and an experimental study of the factors influencing the thermal diffusivity of compacted SYNROC powder is described

  20. Continuous-flow leaching studies of crushed and cored SYNROC

    International Nuclear Information System (INIS)

    Coles, D.G.; Bazan, F.

    1982-01-01

    Both crushed (150- to 300- μm) and cored (1,8- mm-diam) samples of SYNROC have been leached with single-pass continuous-flow leaching equipment. Crushed samples of cesium-hollandite were also leached in a similar experiment. Temperatures used were 25 0 and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC-C were ranked cesium > strontium greater than or equal to calcium > barium > zirconium. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and Pacific Northwest Laboratory 76-68 glass beads. This comparison depends on how the surface areas are determined for each sample. Based on geometric surface areas for SYNROC cores and glass beads, cesium leach rates from SYNROC compare well with both sodium and neptunium leached from the glass. The other elements leached from SYNROC are lower than sodium and neptunium leached from glass. They also vary for each element, while glass shows nearly the same leach rate for both sodium and neptunium

  1. Continuous-flow leaching studies of crushed and cored SYNROC

    International Nuclear Information System (INIS)

    Coles, D.G.; Bazan, F.

    1981-01-01

    Both crushed (150 to 300 μm) and cored (1.8 mm diameter) samples of SYNROC have been leached with the single-pass continuous-flow leaching equipment. Crushed samples of Cs-hollandite were also leached in a similar experiment. Temperatures used were 25 and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC C were ranked Cs > Sr greater than or equal to Ca > Ba > Zr. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and PNL 76-68 glass beads. This comparison depends on how the surface areas are determined for each sample. Based on geometric surface areas for SYNROC cores and glass beads Cs leach rates from SYNROC compare well with both Na and Np leached from the glass. The other elements leached from SYNROC are lower than Na and Np leached from glass. They also vary for each element while glass shows nearly the same leach rate for both Na and Np

  2. Sintering, microstructural and dilatometric studies of combustion synthesized Synroc phases

    International Nuclear Information System (INIS)

    Muthuraman, M.; Patil, K.C.; Senbagaraman, S.; Umarji, A.M.

    1996-01-01

    Sintering, microstructure, and linear thermal expansion properties of Synroc-B and constituent phases, viz. perovskite CaTiO 3 , zirconolite ZrTi 2 O 7 , hollandite (ideal formula BaAl2Ti 6 O 16 ) have been investigated. Synroc-B powder when pelletized and sintered at 1250 C for 2 h achieved >95% theoretical density. Sintered Synroc-B has a linear thermal expansion coefficient α of 8.72 x 10 -6 K -1 and Vicker's microhardness 9.88 GPa. The linear thermal expansion curves did not show any hysteresis indicating the absence of microcracking in the sintered bodies

  3. Continuous-flow leaching studies of crushed and cored SYNROC

    International Nuclear Information System (INIS)

    Coles, D.G.; Bazan, F.

    1980-01-01

    Both crushed (150 to 300 μm) and cored 1.8 mm diameter) samples of SYNROC have been leached with the single-pass continuous-flow leaching equipment. Crushed samples of Cs-hollandite were also leached in a similar experiment. Temperatures used were 25 0 C and 75 0 C and leachates were 0.03 N NaHCO 3 and distilled water. Leaching rates from SYNROC C were ranked Cs > Sr greater than or equal to Ca > Ba > Zr. A comparison of leaching rates is made between crushed SYNROC, cored SYNROC, and PNL 76-68 glass beads. Problems encountered when comparing the leaching rates of different waste forms are discussed

  4. Properties of SYNROC C nuclear-waste form: a state-of-the-art review

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1982-09-01

    SYNROC C is a titanate ceramic waste form designed to contain the waste generated by the reprocessing of commercial nuclear reactor fuel. The properties of SYNROC C are described with particular emphasis on the distribution of chemical elements in SYNROC, the fabrication of good quality specimens, and the chemical durability of SYNROC. Data obtained from testing of natural mineral analogues of SYNROC minerals are briefly discussed. The information available on radiation effects in SYNROC in relation to structural alteration and changes in chemical durability are summarized. 26 references, 2 figures, 18 tables

  5. Mechanical and thermophysical properties of hot-pressed SYNROC B

    International Nuclear Information System (INIS)

    Hoenig, C.L.; Newkirk, H.W.; Otto, R.A.; Brady, R.L.; Brown, A.E.; Ulrich, A.R.; Lum, R.C.

    1981-01-01

    The optimal SYNROC compositons for use with commercial waste are reviewed. Large amounts of powder (about 2.5 kg) were prepared by convention al ceramic operations to test the SYNROC concept on a processing scale. Samples, 15.2 cm in diameter, were hot pressed in graphite, and representative samples were cut for microstructural evaluations. Measured mechanical and thermophysical properties did not vary significantly as a function of sample location and were typical of titanate ceramic materials

  6. Application of SYNROC to high-level defense wastes

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  7. SYNROC process. A geochemical approach to nuclear waste immobilization

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  9. Immobilisation of high level nuclear reactor wastes in SYNROC

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-03-15

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

  10. The preparation of Synroc and its radiation stability

    International Nuclear Information System (INIS)

    Evans, J.P.; Boult, K.A.; Paige, E.L.; Marples, J.A.C.

    1986-12-01

    Samples of Synroc have been made from simulated Highly Active Waste and 'Sandia' precursor supplied by the Australian Atomic Energy Commission, with a view to gaining experience for making fully active samples. The Synroc pellets were of good density (approx. 4.5g.cm -3 ), and and with comparable leach rates to samples made in Australia but the micro-structure was rather coarser. Some Pu-238 doped pellets made previously from low surface area precursors but with the correct phase structure have now received a radiation dose equivalent to an age of 400,000 years for Synroc containing fully active waste. The samples are still intact but their volumes have increased by 6%, in agreement with Australian results on samples damaged by neutron irradiation. (author)

  11. Thermal expansion of U.S. and Australian SYNROC B

    International Nuclear Information System (INIS)

    Kase, H.R.; Case, E.D.; Tesk, J.A.

    1985-01-01

    For the safe disposal of nuclear waste, a synthetic rock (SYNROC) was developed. Continuing research in this field has led to US and Australian versions of SYNROC B. For both materials, the thermal expansion and expansivity have been determined by the temperature range from 296 to 1100 K. Although both versions of SYNROC B have basically the same composition and agree in the major constituent phases, the U.S. version expands slightly more than the Australian one. With increasing temperature, the difference becomes greater and runs up to 3.5% at 1100 K. Because of the good linearity in the temperature dependence of the relative thermal expansion (ΔL/L /sub o/ ), a linear regression was made and the resulting equations determined

  12. Leaching studies on SYNROC at 950C and 2000C

    International Nuclear Information System (INIS)

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

    1982-01-01

    Crushed samples of SYNROC containing 9%, 16% and 20% of simulated high-level nuclear waste were tested for leaching behavior in distilled water at 95 0 C and 200 0 . Leach solutions were analyzed for Cs, Ca, Ba, Sr, Ti, Zr, Nd and U. Results showed that leach rates based on these elements did not change significantly as the waste loading was increased from 9 to 20%. At both temperatures, leach rates showed a decrease as leaching progressed until a plateau level was reached. Plateau leach rates, which were between 10 and 100 times lower than initial leach rates, reflect the expected long term leaching behaviour of the samples. Plateau values of leach rates for SYNROC depend on the element being leached. Highest values are found for Cs and Ba (1 to 2 x 10 -7 g/cm 2 d at 95 0 C) and lowest values for U (5 x 10 -10 g/cm 2 d at 95 0 C). Increasing leaching temperature to 200 0 C produces higher leach rates for all elements except Nd. Comparison of SYNROC leach rate data with that for PNL 76-68 glass shows that at 200 0 C the leach rate for U from SYNROC is 3000 times less than that from glass. (Auth.)

  13. Operating procedures for the manufacture of radioactive SYNROC in the actinide laboratory

    International Nuclear Information System (INIS)

    Western, K.F.

    1986-03-01

    The purpose of this manual is to acquaint the operator with the procedures required to manufacture SYNROC-containing radioactive materials in the SYNROC actinide laboratory, Lucas Heights Research Laboratories. The actinide-doped SYNROC production facility is a series of four interconnected glove boxes and one free-standing glove box. The samples of radioactive SYNROC produced in the actinide laboratory are used to carry out physical testing of the product at various laboratories on site, e.g. leach testing, auto-radiographic examination, electron-microscopc examination, atomic absorption spectrophotometry and analysis

  14. Thermal durability of modified Synroc material as reactor fuel matrix

    International Nuclear Information System (INIS)

    Kikuchi, Akira; Kanazawa, Hiroyuki; Togashi, Yoshihiro; Matumoto, Seiichiro; Nishino, Yasuharu; Ohwada, Isao; Nakata, Masahito; Amano, Hidetoshi; Mitamura, Hisayoshi

    1994-08-01

    A Synroc, a polyphase titanate ceramics composed of three mineral phases (perovskite, hollandite and zirconolite), has an excellent performance of immobilization of high level nuclear waste. A working group in the Department of Hot Laboratories paid special attention to this merit and started a development study on a LWR fuel named 'Waste Disposal Possible (WDP) Fuel', which has the two functions of a reactor fuel and a waste form. The present paper mainly describes thermal durability of a modified Synroc material, which is essentially important for applying the material to a fuel matrix. The two kinds of Synroc specimens, designated 'SM' as modified and 'SB' as a reference, were prepared by hot-pressing and annealed at 1200degC to 1500degC for 30 min in air. Unexpected and peculiar spherical voids were observed in the specimen SM at 1400degC and 1500degC, which caused the specimen swelling. The formation of the voids depends significantly on the existence of spherical precipitates seen in the as-fabricated specimen including latent micropores with high pressure. On the other hand, the heat treatment at 1500degC formed additional new phases, designated 'Phase A' for the specimen SB and 'Phase X' for SM. Phase A is a decomposition product of hollandite and Phase X a reaction product of Phase A and perovskite in the spherical voids. Furthermore, additional information and thermal properties examined are presented in Appendix 1 and Appendix 2, respectively. It was recognized that the modified Synroc specimen SM had excellent thermal properties. (author)

  15. Experience gained with the Synroc demonstration plant at ANSTO and its relevance to plutonium immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Jostsons, A.; Ridal, A.; Mercer, D.J.; Vance, E.R.L. [Australian Nuclear Science and Technology Organisation, Menai (Australia)

    1996-05-01

    The Synroc Demonstration Plant (SDP) was designed and constructed at Lucas Heights to demonstrate the feasibility of Synroc production on a commercial scale (10 kg/hr) with simulated Purex liquid HLW. Since commissioning of the SDP in 1987, over 6000 kg of Synroc has been fabricated with a range of feeds and waste loadings. The SDP utilises uniaxial hot-pressing to consolidate Synroc. Pressureless sintering and hot-isostatic pressing have also been studied at smaller scales. The results of this extensive process development have been incorporated in a conceptual design for a radioactive plant to condition HLW from a reprocessing plant with a capacity to treat 800 tpa of spent LWR fuel. Synroic containing TRU, including Pu, and fission products has been fabricated and characterised in a glove-box facility and hot cells, respectively. The extensive experience in processing of Synroc over the past 15 years is summarised and its relevance to immobilization of surplus plutonium is discussed.

  16. Evaluation of critical properties of SYNROC for disposal of high level radioactive waste

    International Nuclear Information System (INIS)

    Segall, R.L.; Myhra, S.; Smart, R.; Turner, P.S.

    1983-12-01

    Factors influencing leaching of caesium and strontium and the contribution to improved leach resistance of SYNROC C fabrication in the Australian Atomic Energy Commission SYNROC program are examined. The results confirm the good caesium leach resistance under hydrothermal conditions of properly fabricated SYNROC and provide some insight into mechanisms that control the higher initial (first day) leach rates. Studies have been completed detailing the partitioning of different elements into the SYNROC phases. The microstructure of SYNROC C has been extensively investigated and correlated with leach resistance. Attack at specific regions in the surface and changes in the composition of the surface region have been monitored with a variety of electron microscopic and surface analytical techniques. A set of mechanisms for the leaching and dissolution of SYNROC are proposed. On the basis of these different lines of research, recommendation of optimal conditions for fabrication are given. This has resulted in the manufacture of high quality SYNROC on a routine basis. Other aspects of quality control particularly of variations in waste loading and large-scale operation, are also reported

  17. Phase equilibria, leaching characteristics and ceramic processing of SYNROC D formulations for US defense wastes

    International Nuclear Information System (INIS)

    Newkirk, H.; Ryerson, F.; Coles, D.; Hoenig, C.; Rozsa, R.; Rossington, C.; Bazan, F.; Tewhey, J.

    1980-01-01

    The assemblage of coexisting phases in SYNROC D is perovskite, zirconolite, nepheline and spinel. Cesium from the supernate is to be immobilized in hollandite. In the current processing scheme, presynthesized granules of hollandite are added to calcined SYNROC D powders prior to hot procesing or sintering. The disposition of inert and radwaste components of Savannah River Plant (SRP) wastes in SYNROC D formulations has been determined by means of optical microscopy, XRD, XRF, SEM, STEM, electron microprobe analysis and autoradiography. A summary of results is presented. Leaching studies of SYNROC D have been done by means of static, high temperature experiments and continuous-flow experiments. The data reported are from high-temperature experiments (distilled water, powdered sample, 150 0 C, one day). The elements reported are the only ones observed in the leachate. Analysis was done by means of XRF. The flowsheet which depicts the current experimental methods that are being employed at LLNL to produce SYNROC D samples containing presynthesized Cs-bearing hollandite is presented. The starting material for SYNROC D (high Fe, high Al and composite compositions) is simulated sludge obtained in 55 gallon quantities from Southwestern Chemical Corporation. Hot pressing temperatures for SYNROC D are 1000 to 1150 0 C. Hot pressing temperatures for hollandite are 1200 to 1400 0 C

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  19. Immobilization of sodium and phosphorus-bearing PW-7a waste in SYNROC. Progress report

    International Nuclear Information System (INIS)

    Ringwood, A.E.

    1982-01-01

    The phosphorus, sodium and gadolinium-rich PW-7a waste can be successfully incorporated in SYNROC-C. However, a new accessory phase, a Ca,Na,Ba phosphate isostructural with Ca 5 Na 2 (PO 4 ) 4 apppears in the SYNROC mineralogy. There is no evidence for the partition of key radionuclides (e.g. Sr, REE and hence actinides) into this phosphate. Its poor resistance to groundwater dissolution, whilst hardly desirable, may therefore not have a serious effect on the leaching performance of SYNROC containing PW-7a. 9 tables

  20. A state-of-the-art report on the formulation and characterization of synroc

    International Nuclear Information System (INIS)

    Jung, C. H.; Park, J. Y.; Oh, S. J.; Kim, H. Y.; Kim, I. T.

    1998-09-01

    Synroc (Synthetic Rock), a titanate-based ceramic originally proposed by Prof. A. Ringwood (ANU) and designed for the immobilization of high level nuclear waste (HLW), consists of three principal phases such as hollandite, zirconolite and perovskite. Nearly all the fission products and actinides in HLW can be incorporated as solid-solution in at least one of these phase. The preferred form of Synroc can be obtained up to 20 % of high level waste calcine to form dilute solid solution. The constituent minerals, or close structural analogues, have survived in a wide range of geochemical environments for periods of 20-2000 Myr while immobilizing the same elements present in nuclear waste. A dense, compact, and mechanically strong form of Synroc can be formed by hot pressing reactive precursor powders at about 1200 dg C and 20 MPa. In this state-of-the-art report, formulation method and characterization of Syroc with respect to the crystal structure, the consisting substances, types, etc. were reviewed. Additionally, a new promising powder process, C ombustion Process , was proposed and the properties of the combustion-synthesized powder were described. An international cooperative program between JAERI and ANSTO, and US patents for early Synroc research in Australia were also introduced. From the literatures review, Synroc is expected to have advantages in using as an immobilizer of HLW. Therefore, a systematic research to develop the Synroc is needed. (author). 53 refs., 2 tabs., 16 figs

  1. Leach testing of SYNROC and glass samples at 85 and 200/degree/C

    International Nuclear Information System (INIS)

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

    1981-01-01

    Leach tests were conducted on 0.5 g disc samples of SYNROC and two glass types using distilled water at 85 and 200/degree/C. No leaching was detected for SYNROC at either temperature. Thus, the upper limit on leach rate for SYNROC is <0.005 g/m/sup 2/d. Waste glass PNL 76-68 had leach rates of 1.4 g/m/sup 2/ d at 85/degree/C and 8.9 g/m/sup 2/ d at 200/degree/C, while 73-1 glass frit had a leach rate of 41 g/m/sup 2/ d at 200/degree/C. The leach tests were repeated in the presence of rock powders. Again, no leaching was measurable for SYNROC. PNL 76-68 glass had leach rates between 4 and 23 g/m/sup 2/ d at 200/degree/C and 73-1 frit leached at rates between 29 and 176 g/m/sup 2/ d at 200/degree/C. Tests were also conducted on crushed glass samples (PNL 76-68, 100-200 /mu/m size fraction). Bulk leach rates were calculated based on measurement of Ca, Cs, and U in the leach solutions. The results of the leach tests show that SYNROC is several orders of magnitude more resistant to leaching than glass

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

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

  3. Preparation and properties of SYNROC D containing simulated Savannah River Plant high-level defense waste

    International Nuclear Information System (INIS)

    Hoenig, C.; Rozsa, R.; Bazan, F.; Otto, R.; Grens, J.

    1981-01-01

    We describe in detail the formulation and processing steps used to prepare all SYNROC D samples tested in the Comparative Leach Testing Program at the Savannah River Laboratory. We also discuss how the composition of the Savannah River Plant sludge influences the formulation and ultimate preparation of SYNROC D. Mechanical properties are reported in the categories of elastic constants, flexural and compressive strengths, and microhardness; thermal expansion and thermal conductivity results are presented. The thermal expansion data indicated the presence of significant residual strain and the possibility of an unidentified amorphous or glassy phase in the microstructure. We summarize the standardized (MCC) leaching results for both crushed Synroc and monoliths in deionized water, silicate water, and salt brine at 90 0 C and 150 0 C

  4. Preparation and properties of SYNROC D containing simulated Savannah River Plant high-level defense waste

    Energy Technology Data Exchange (ETDEWEB)

    Hoenig, C.; Rozsa, R.; Bazan, F.; Otto, R.; Grens, J.

    1981-07-23

    We describe in detail the formulation and processing steps used to prepare all SYNROC D samples tested in the Comparative Leach Testing Program at the Savannah River Laboratory. We also discuss how the composition of the Savannah River Plant sludge influences the formulation and ultimate preparation of SYNROC D. Mechanical properties are reported in the categories of elastic constants, flexural and compressive strengths, and microhardness; thermal expansion and thermal conductivity results are presented. The thermal expansion data indicated the presence of significant residual strain and the possibility of an unidentified amorphous or glassy phase in the microstructure. We summarize the standardized (MCC) leaching results for both crushed Synroc and monoliths in deionized water, silicate water, and salt brine at 90/sup 0/C and 150/sup 0/C.

  5. Formulation of SYNROC-D additives for Savannah River Plant high-level radioactive waste

    International Nuclear Information System (INIS)

    Ryerson, F.J.; Burr, K.; Rozsa, R.

    1981-12-01

    SYNROC-D is a multiphase ceramic waste form consisting of nepheline, zirconolite, perovskite, and spinel. It has been formulated for the immobilization of high-level radioactive wastes now stored at Savannah River Plant (SRP) near Aiken, South Carolina. This report utilizes existing experimental data to develop a method for calculating additives to these waste products. This method calculates additions based on variations of mineral compositions as a function of sludge composition and radionuclide partitioning among the SYNROC phases. Based on these calculations, a FORTRAN program called ADSYN has been developed to determine the proper reagent proportions to be added to the SRP sludges

  6. Densification studies of Synroc D for high-level defense waste

    International Nuclear Information System (INIS)

    Hoenig, C.; Otto, R.; Campbell, J.

    1983-01-01

    Small- to medium-scale densification experiments were conducted on Synroc D using graphite dies and metal canisters. Pressures at elevated temperatures were applied both isostatically (HIP) and unidirectionally (HUP). Spray-dried/calcined powders formulated for composite or average sludge compositions exhibited initial packing densities of about 25% theoretical. Final densities were in the range of 95 to 99% theoretical, depending on applied pressure and temperature. In final-stage HUP densification, we have found that porosity varies exponentially with time acording to the well-known expression P + P 0 exp(-K 0 t). The rate constant (K 0 ) has the Arrhenius form K 0 = Asigma exp(-E/RT) which includes a stress or pressure term. Rate constants are calculated from approximately 20 densification experiments conducted under a wide range of conditions; activation energies in the range of 20 to 35 kcal/mole were calculated for the densification process. HIP densification and leaching results are reported for experiments with a wide range of variables: pressure (3 to 30 ksi), temperature (900 to 1200 0 C), redox calcination method, powder fill density and metal canister material. The results support the conclusion that HUP and HIP densification parameters are very similar and that Synroc-D leaching behavior is essentially independent of density in the range of 90 to 100% theoretical.The densification of Synroc D in a collapsible metal-bellows canister has been simulated by means of modeling calculations. Radial buckling tendencies were also evaluated. Results from large-scale HIP experiments are also reported. Up to 50 kg of Synroc D was densified to greater than 99% theoretical density in a metal-bellows canister 36 cm diameter by 24 cm in height. These data were used as a guide to make recommendations for the full-scale HIP densification of Synroc D using metal-bellows canisters

  7. Final report on fabrication and study of SYNROC containing radioactive waste elements

    International Nuclear Information System (INIS)

    Reeve, K.D.; Levins, D.M.; Seatonberry, B.W.; Ryan, R.K.; Hart, K.P.; Stevens, G.T.

    1987-01-01

    Two facilities for the fabrication and testing of Synroc samples containing separate additions of the transuranic actinides americium, plutonium, curium and neptunium, a fission product solution, and two radioisotopes of caesium and strontium were designed, built and operated by the AAEC at the Lucas Heights Research Laboratories. Twenty-one 75 g batches of radioactive Synroc were made and representative samples were characterised by alpha track etching, scanning electron microscopy and aqueous leach testing, mostly at 70 deg C. Where comparisons were possible, radioactive fission products behaved as expected from non-radioactive tests. The leaching behaviour of the actinides was complex but as a group they were the least leachable of all the elements studied

  8. Performance of borosilicate glass, Synroc and spent fuel as nuclear waste forms

    International Nuclear Information System (INIS)

    Lutze, W.; Grambow, B.; Ewing, R.C.

    1990-01-01

    Presently, there are three prominent waste forms under consideration for the disposal of high-level waste: Borosilicate glass and Synroc for high-level radioactive waste from fuel reprocessing and spent fuel as the waste form for non-reprocessed fuel. Using the present experimental data base, one may compare the performance of these three waste forms under repository relevant conditions. In low water flow regimes and at temperatures less than 100 degree C, the fractional release rates of all three waste forms are low, on the order of 10-7/d or less and may decrease with time. Under these conditions the three waste forms behave similarly. At elevated temperatures or in high flow regimes, the durability of borosilicate glass will be much less than that of Synroc, and thus, for certain disposal schemes (e.g., deep burial) Synroc is preferable. All predictions of the long-term behavior are based on the extrapolation of short term experimental data, we point out that appropriate and useful natural analogues are available for each of these waste forms and should be used in the performance assessment of each waste form's long-term behavior. 14 refs

  9. Incorporation of high-level wastes in SYNROC: results from recent process-engineering studies at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    Campbell, J.H.; Hoenig, C.L.; Ackerman, F.J.; Peters, P.E.; Grens, J.Z.

    1982-01-01

    In this paper, highlights from recent engineering research and development, in particular, results from fluidized bed calcination studies of SYNROC slurry are summarized. A schematic diagram of the envisioned SYNROC process (at this stage of development) is also presented. It shows the use of a fluidized bed calciner to prepare SYNROC powder that is then fed to a storage hopper. Bellows-type canisters are filled, evacuated, sealed and preheated. The preheated canisters are loaded into a hot isotactic pressing unit where they are densified, then removed and cooled and finally loaded into a waste storage container. After sealing, this container is decontaminated and transferred to the interim storage facility and then, ultimately, to an underground repository

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

    International Nuclear Information System (INIS)

    Chen, C.

    1982-01-01

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

  11. Comparison of the properties of simulated synroc synthesized by sol-gel and a novel co - precipitation method

    International Nuclear Information System (INIS)

    Potdar, H.S.; Vijayanand, S.; Khaja Mohaideen, K.; Joy, P.A.; Raja Madhavan, R.; Kutty, K.V.G.; Ambashta, R.D.; Wattal, P.K.

    2009-01-01

    Synroc is a multiphase dense titanate based ceramic designed for the incorporation of high-level waste (HLW) from the reprocessing of spent nuclear fuel. Synroc or synthetic rock consists of four main titanate phases - zirconolite (CaZrTi 2 O 7 ), hollandite (BaAlO 2 Ti 6 O 16 ), perovskite (CaTiO 3 ) and rutile (TiO 2 ), with the matrix composition as shown in Table 1. It is known that these phases have the capacity to incorporate most of the elements into their crystal structures which are present in the HLW derived from the reprocessing of spent nuclear fuel from power reactors. Synroc is considered as the most effective and durable means of immobilising various forms of high-level radioactive wastes for disposal. Synroc is also considered as a low-risk, tailored waste form, offering higher waste loading and over all cost savings. Simulated synroc precursor powders are typically produced by advanced wet chemical methods such as alkoxide hydrolysis and sol-gel routes. These routes were developed to produce powders with well defined physical and chemical characteristics such as correct chemical composition, high degree of homogeneity, reactivity and readily densifiable material to 99% of theoretical density during hot isostatic pressing. However, the reported alkoxide hydrolysis and hydroxide routes suffer from several disadvantages such as use of large quantities of organic solvents and their disposal as effluent, difficulty in maintaining exact chemical composition, use of costly alkoxide precursors which are moisture sensitive and require critical processing conditions to control their rate of hydrolysis, etc. In the present work we report a comparative study the characteristics of synroc-C (14% waste loading) powders and sintered pellets synthesized by the known alkoxide hydrolysis method and a simple chemical co-precipitation route developed by us. The advantages of the co-precipitation route are its simplicity, ease of handling and utilization of cheaper raw

  12. Analytical methods for the determintion of some elements and Fe+2 to Fe+3 ratio in simulated sludges and Synroc formulations

    International Nuclear Information System (INIS)

    Lim, R.

    1981-10-01

    Analytical methods for the determination of Fe, Al, Mn, Ca, Ni, Na, Sr, Cs, Ti, and Ba in simulated sludges and Synroc formulations are discussed. These are the elements that may be completed by atomic absorption spectroscopy, AAS. AAS methods are complicated by the dissolution methods used. These problems are discussed. In addition, the method used for the determination of Fe +2 to Fe +3 ratio is presented

  13. Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

    Energy Technology Data Exchange (ETDEWEB)

    Maio, Vince [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

  14. Small angle X-ray scattering by TiO2/ZrO2 mixed oxide particles and a Synroc precursor

    International Nuclear Information System (INIS)

    Gazeau, D.; Zemb, T.; Amal, R.; Bartlett, J.

    1992-09-01

    This high resolution small angle X-ray scattering study of a concentrated oxide sol, precursor of the SYNROC matrix for the storage of the high level radioactive waste, evidences a locally cylindrical microstructure. Locally, nanometric cylinders show disordered axis with some concentration dependent connections. This microstructure explains the paradoxal stability of this oxide dispersions upon the addition of concentrated acidic solutions. This stability has a steric origin and electrostatic repulsions are not needed. The addition of aluminium to the initial titanium-zirconium mixture enhances branching on the locally cylindrical microstructure. Finally, we show that the solid powder obtained after calcination (drying) of the sol has the same specific area (∼ 1000 m 2 /g) than the sol. (Author). 23 refs., 7 figs., 1 tab

  15. Characterization of a Pu-bearing zirconolite-rich synroc

    International Nuclear Information System (INIS)

    Buck, E.C.; Ebbinghaus, B.; Bakel, A.J.; Bates, J.K.

    1996-01-01

    A titanate-based ceramic waste form, rich in phases structurally related to zirconolite (CaZrTi 2 O 7 ), is being developed as a possible method for immobilizing excess plutonium from dismantled nuclear weapons. As part of this program, Lawrence Livermore National Laboratory (LLNL) produced several ceramics that were then characterized at Argonne National Laboratory (ANL). The plutonium- loaded ceramic was found to contain a Pu-Gd zirconolite phase but also contained plutonium titanates, Gd-polymignyte, and a series of other phases. In addition, much of the Pu was remained as PuO 2- x . The Pu oxidation state in the zirconolite was determined to be mainly Pu 4+ , although some Pu 3+ was believed to be present

  16. Naturally-occurring zirconolites - analogues for the long-term encapsulation of actinides in synroc

    International Nuclear Information System (INIS)

    Hart, K.P.; Lumpkin, G.R.; Giere, R.; Williams, C.T.; McGlinn, P.J.; Payne, T.E.

    1996-01-01

    The use of natural zirconolites to assess the effect of α-decay damage and geochemical alteration on the release of actinides from HLW wasteforms is critically examined. There is evidence that the natural zirconolites provide a good chemical and radiation damage analogy for the HLW wasteforms, but additional work is required to define the geochemical environments in which zirconolite is stable or unstable (e.g., suffering corrosion or chemical alteration, including loss of actinides). (orig.)

  17. 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

  18. Benefits of nuclear reactor still unclear

    International Nuclear Information System (INIS)

    Allen, Barry

    1997-01-01

    The author questions the Australian Government decision to build a new reactor at Lucas Heights and to reject the proposal for a nuclear waste reprocessing and disposal using Australia's Synroc technology. He argued that Australia should have looked to the future(Synroc) instead of investing in dated technology (Reactor) and sees Synroc technology having much more potential to generate foreign currency if the increasing need for waste disposal facilities in the region are considered

  19. Progress report - August 1991

    International Nuclear Information System (INIS)

    1991-08-01

    This report has been prepared by the Synroc Study Group (SSG), comprising staff members of The Australian Nuclear Science and Technology Organisation, The Australian National University, The Broken Hill Proprietary Company Limited, CRA Limited, Energy Resources of Australia Limited and Western Mining Corporation Limited. It also draws upon work undertaken for the SSG by consultants from Nuclear Assurance Corporation (market estimates), the Swedish Nuclear Fuel and Waste Management Company (SKB) (cost estimates and other data) and Wave Hill Associates (US and international perspectives). Synroc is applicable solely to the immobilisation of liquid high-level waste after such waste has been separated from spent fuel in a reprocessing operation; use of Synroc therefore requires prior construction of a reprocessing plant. The study identifies five broad options in which Synroc can contribute to the safe disposal of nuclear spent fuel. These are: licensing Synroc for overseas organisations - allowing overseas use of the technology with royalties remitted to Australia; participation in overseas Synroc plants - using Australian expertise to trial Synroc facilities; reprocessing and Synroc operations in Australia with waste re-exported to customers for disposal overseas; establishment of an integrated spent fuel management facility outside Australia including a Synroc facility and final disposal; establish an integrated spent fuel management facility in Australia, including a Synroc facility and final disposal. An account of the SSG's public acceptance initiatives and activities to date, is found in Appendix II. Appendix III contains material derived from the SKB work on the costs of providing various waste management services, and outlines the economic factors affecting estimates of these costs. 86 refs., 13 tabs., 9 figs

  20. Safe immobilization of high-level nuclear reactor wastes

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  1. Research On Stabilization Of Radioactive Waste By Method Of SYNROCK Ceramic

    International Nuclear Information System (INIS)

    Nguyen Hoang Lan; Nguyen Ba Tien; Vuong Huu Anh; Nguyen An Thai

    2014-01-01

    Separate phases from SYNROC polyphases ceramic were investigated to fabricate completely SYNROC and the distribution of stable isotopes (Sr) in SYNROC matrix was surveyed simultaneously with leaching test. The experimental conditions: 13.5 x 11mm pressed pellet SYNROC with pressure of 2.5 - 3 tons/cm 2 , sintering temperature t tk = 1250 o C, thermal lifting velocity v t = 20 o C/min with 2 hours prolongation in 1250 o C, Sr loading amount was 7% mole, the results showed that pellets contain 3 phases perovskite CaTiO 3 , zirconolite CaZrTi 2 O 7 , hollandite BaAl 2 Ti 6 O 16 with average density of 4.1 g/cm 3 , leaching rate R (g/m 2 .d) of 10 -6 , 10 -5 for Ti, Sr respectively. (author)

  2. Alternative-waste-form evaluation for Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Gould, T.H. Jr.; Crandall, J.L.

    1982-01-01

    Results of the waste form evaluation are summarized as: risks of human exposure are comparable and extremely small for either borosilicate glass or Synroc ceramic. Waste form properties are more than adequate for either form. The waste form decision can therefore be made on the basis of practicality and cost effectiveness. Synroc offers lower costs for transportation and emplacement. The borosilicate glass form offers the lowest total disposal cost, much simpler and less costly production, an established and proven process, lower future development costs, and an earlier startup of the DWPF

  3. Deep repositories for waste central to uranium debate

    International Nuclear Information System (INIS)

    Kannegieter, T.

    1991-01-01

    While no deep repositories for high level wastes (HLW) have yet been constructed it is shown that technology to safely entomb the wastes for tens of thousands of years already exists. The borosilicate glass (vitrification) developed in France has been accepted by all countries who are reprocessing. Meanwhile, the Australian Synroc has not yet been put into service. Synroc developers at the Australian Nuclear Science and Technology Organization believe it will be the second generation waste form. The advantages and disadvantages of both technologies are briefly discussed as well as some of the regulatory, political, legal and technical conflicts surrounding the issue of HLW repositories. 1 tab., ills

  4. Solid state synthesis and structural refinement of polycrystalline La ...

    Indian Academy of Sciences (India)

    Perovskite structure based ceramic precursors have a characteristic property of substitution in the ``A" site of the ABO3 structure. This makes them a potential material for nuclear waste management in synthetic rock (SYNROC) technology. In order to simulate the mechanism of rare earth fixation in perovskite, La ...

  5. Synthesis of CaTiO 3 from calcium titanyl oxalate hexahydrate (CTO)

    Indian Academy of Sciences (India)

    Calcium titanate, CaTiO3, an important microwave dielectric material and one of major phases in synroc (synthetic rock), a titanate ceramic with potential application for fixation of high level nuclear waste was synthesized from calcium titanyl oxalate [CaTiO (C2O4)2.6H2O] (CTO) by employing microwave heating technique.

  6. Energy and technology review

    International Nuclear Information System (INIS)

    Stowers, I.F.; Crawford, R.B.; Esser, M.A.; O'Neal, E.

    1981-12-01

    Research programs at Lawrence Livermore Laboratories are described. These include: the generation of intense electron beams for military applications; SYNROC, a permanent means of radioactive-waste storage in synthetic rock compounds; and studies of respiration using a positron camera with radioisotopes produced in the 100-MeV electron linear accelerator

  7. Australia's uranium and the international nuclear industry

    International Nuclear Information System (INIS)

    Ringwood, A.E.

    1983-11-01

    A disposal strategy for high-level radioactive waste is presented. The waste is incorporated in SYNROC which is then buried in deep drill holes in a stable geological environment. It is suggested that acceptance of the safety of this strategy would remove a primary objection to the mining of Australian uranium. Further Australian involvement in the fuel cycle is advocated

  8. Safety in depth for nuclear waste disposal

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-11-27

    A nuclear waste disposal strategy is described in which the radionuclides are immobilised in widely-dispersed drill holes in an extremely stable and leach resistant titanate ceramic form (SYNROC) at depths of 1500 to 4000 metres. The advantages of this method over that of burying such wastes in large centralised mined repositories at 500 to 700 metres in suitable geological strata are examined.

  9. Assessment of methods for immobilizing reprocessed radioactive waste

    Science.gov (United States)

    Murthy, M. K.; Baranyi, A. D.

    1980-05-01

    Nuclear waste forms presently used for the disposal of high level wastes and other potential waste forms under development were studied. The following waste forms were considered: Borosilicate glass, high silica glass, glassceramics, supercalcine ceramics, synroc ceramics, borosilicate glass beads in a metal matrix, supercalcine and synroc ceramics in a metal matrix and coated ceramics. The best developed wasteform, both in terms of overall product quality and process development, is monolithic borosilicate glass. However, hydrothermal instability is a major concern. Borosilicate glass in metal matrix waste form has better properties than monolithic borosilicate glass waste form. The process was proven on a pilot scale. Hence, it is considered very close to monolithic glass in terms of overall development. The product qualities of the other waste forms are better than borosilicate glass. However, process development for these alternative waste forms is still in a conceptual stage.

  10. The use of metal alkoxides in the preparation of ceramic powders

    International Nuclear Information System (INIS)

    Chetcuti, A.M.; Woolfrey, J.L.

    1982-01-01

    The production of fine, chemically homogeneous and highly reactive powder is particularly desirable where the synthesis and fabrication of multicomponent ceramic systems, such as SYNROC, are concerned. To produce good sinterable material, a preparation technique that allows intimate mixing of all reacting species is desirable. Traditional routes for preparing fine powders have involved ball-milling metal oxides and spray-drying or flash-drying the resulting oxide slurries. The hydrolysis of metal alkoxides has been investigated as a technique to produce fine powders. The preparation of SYNROC B powder from alkoxides involves hydrolysing a mixture of titanium and zirconium alkoxides. The precipitated product is then blended with Al 3 + , Ba 2 + and Ca 2 + nitrate solution

  11. Current ANSTO research on wasteform development

    Energy Technology Data Exchange (ETDEWEB)

    Vance, E R; Begg, B D; Stewart, M W.A.; Moricca, S; Smith, K L; Walls, P A; Perera, D S; Day, R A; Carter, M L; McGlinn, P J; Zhang, Y; Thomas, B [Australian Nuclear Science and Technology Organisation, Menai, NSW (Australia). Materials and Engineering Science

    2003-07-01

    In 1978, Ringwood suggested ceramic assemblages of titanate minerals could be used to incorporate high-level waste from nuclear fuel reprocessing. In these assemblages waste ions are dilutely incorporated into the crystalline mineral-analogue phases. Synroc-C is one of the early titanate assemblages and it has become the archetype from which waste forms for various applications have been derived. Table 1 shows the phase constitution of synroc-C, containing 20 wt% HLW, and the radionuclides which can be incorporated in the various phases. This material was consolidated into a dense ceramic by uniaxial hot pressing at {approx} 1150 deg C. ANSTO has subsequently undertaken both contract and collaborative work on a variety of waste streams that are briefly described as well as extensive range of wasteform characterisation.

  12. Thermodynamic stability and kinetics of perovskite dissolution

    Energy Technology Data Exchange (ETDEWEB)

    Nesbitt, H W; Bancroft, G M; Fyfe, W S; Karkhanis, S N; Nishijima, A [Western Ontario Univ., London (Canada); Shin, S [National Chemical Lab. for Industry, Tsukuba (Japan)

    1981-01-29

    Perovskite, a SYNROC host mineral for nuclear wastes, is thermodynamically unstable in natural waters and in association with common minerals. Leach experiments demonstrate that CaTiO/sub 3/ (perovskite), SrTiO/sub 3/ and BaTiO/sub 3/ are as reactive as some silicate glasses below 100/sup 0/C, but leach much more slowly than glasses above 100/sup 0/C.

  13. Crystalline matter for solidification of highly radioactive wastes

    International Nuclear Information System (INIS)

    Grauer, R.

    1984-02-01

    Highly active wastes from reprocessed nuclear fuels must be incorporated into a solid chemically resistant inorganic matrix prior to final storage. One possible alternative to glassification is to embed the complex oxide mixture in a crystalline ceramic. A discussion from the structural and chemical viewpoint is presented giving guidelines for the selection and development of such a product. The chemical and phase composition concerning the most important developments are described. SYNROC is the most highly developed solid ceramic that has been evaluated to date for power reactor wastes. However, its testing and development so far has been restricted to simulated inactive materials. One of the most important aspects of solid high activity wastes is their behaviour in water. SYNROC reacts more slowly than glasses with water at temperatures over 100 0 C. Its low release of actinides under these conditions is remarkable. At temperatures under 100 0 C the important nuclide Cs 137 is released from SYNROC and from glasses at comparable rates. These assertions concerning chemical stability are however based on short term experiments, which have not considered the possibly complex interactions occurring during final storage. The information is therefore insufficient to describe the basic model required to predict long term behaviour under final storage conditions. Finally the report makes recommendations for a further programme of work. (Auth.)

  14. Thermodynamic stability and kinetic dissolution of perovskite in natural waters

    International Nuclear Information System (INIS)

    Nesbitt, H.W.; Bancroft, G.M.; Fyfe, W.S.; Karkhanis, S.; Melling, P.; Nishijima, A.

    1981-01-01

    Ringwood and coworkers have recently proposed using titanates and zirconates as hosts for nuclear waste in the Synroc B process. Three minerals are used as hosts: perovskite (CaTiO 3 ), Ba-hollandite (BaAl 2 Ti 6 O 16 ), and zirconolite (CaZrTi 2 O 7 ). The Synroc philosophy relies heavily on geological and geochemical observations in selecting stable host minerals. Although it has been recognized that the Synroc minerals are not thermodynamically compatible with siliceous rocks, the minerals are considered to be thermodynamically stable in the presence of water, and it has been reported that these minerals are kinetically stable under high-temperature (up to 900 0 C) hydrothermal conditions. Detailed thermodynamic calculations and leach tests have been performed which demonstrate: first, that perovskite is thermodynamically unstable in all known natural waters; and second, that pervoskite leaches at a significant rate even at 100 0 C. Hydrothermal leach tests have been made on natural and synthetic perovskite and perovskite analogues between 100 0 C and 300 0 C. Weight losses and solution concentrations were monitored. The results reported previously in the literature also show that perovskite is kinetically unstable in the presence of common silicates. Our results show that perovskite may be no more stable than siliceous glasses, such as rhyolite, which have been studied previously. Geologic evidence from common alkaline rocks also indicates that hollandite and zirconolite probably will not survive in common rock matrices

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-08-01

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

  16. The CaO-TiO2-ZrO2 system at 1,200 degree C and the solubilities of Hf and Gd in zirconolite

    International Nuclear Information System (INIS)

    Swenson, D.; Nieh, T.G.; Fournelle, J.H.

    1995-12-01

    In recent years, significant technological advancements have been made in the Synroc scheme for the immobilization high-level nuclear waste. However, many basic scientific issues related to Synroc fabrication have yet to be addressed. The CaO-TiO 2 -ZrO 2 system is an integral part of the Synroc formulation. Phase equilibria are established in the CaO-TiO 2 -ZrO 2 system at 1,200 C, using X-ray diffraction and electron probe microanalysis. The existence of two previously reported ternary phases, zirconolite (CaZrTi 2 O 7 ) and calzirtite (Ca 2 Zr 5 Ti 2 O 16 ), is confirmed. Each of these phases exhibits a significant range of homogeneity between TiO 2 and ZrO 2 while maintaining a nearly constant concentration of CaO. The ternary solubilities of the constituent binary phases are found to be negligible, with the exceptions of the perovskites, which display mutual solubility of at least 22 mol.% and may in fact form a series of continuous solid solutions. The solubilities of Hf and Gd in zirconolite are also investigated. While Hf-bearing samples did not reach thermodynamic equilibrium under the experimental conditions employed, the existence of a Hf analog to zirconolite, CaHfTi 2 O 7 , is conclusively demonstrated. The phase is stable at the stoichiometric composition, and its lattice parameters are very close to those reported in the literature for stoichiometric zirconolite. A Gd-bearing sample of the composition Ca 0.88 Zr 0.88 Gd 9.24 Ti 2 O 7 is found to be essentially single phase zirconolite, in agreement with previous investigations at higher temperatures

  17. Processability analysis of candidate waste forms

    International Nuclear Information System (INIS)

    Gould, T.H. Jr.; Dunson, J.B. Jr.; Eisenberg, A.M.; Haight, H.G. Jr.; Mello, V.E.; Schuyler, R.L. III.

    1982-01-01

    A quantitative merit evaluation, or processability analysis, was performed to assess the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste form candidates. The reference borosilicate glass process was rated as the simplest, followed by FUETAP concrete, glass marbles in a lead matrix, high-silica glass, crystalline ceramics (SYNROC-D and tailored ceramics), and coated ceramic particles. Cost estimates for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities are also reported

  18. Thirty-first annual report 1982-83

    International Nuclear Information System (INIS)

    1983-01-01

    Activities and research at the Australian Atomic Energy Commission are reported. The research program is divided into five fields: nuclear technology, the nuclear fuel cycle, environmental science, applications of radioisotopes and radiation, and nuclear science. Within these five areas details are given of the fusion research program, a small effort on fission, developmental work on the SYNROC concept for the immobilisation of high-level waste, studies related to the environmental effects of uranium mining, work in isotope hydrology, radiopharmaceutical research, investigations into the irradiation of foods and medical products, industrial applications of radioisotopes and radiations, nuclear physics research and neutron scattering studies

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

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.

    1981-01-01

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

  20. Assessment of processes, facilities, and costs for alternative solid forms for immobilization of SRP defense waste

    International Nuclear Information System (INIS)

    Dunson, J.B. Jr.; Eisenberg, A.M.; Schuyler, R.L. III; Haight, H.G. Jr.; Mello, V.E.; Gould, T.H. Jr.; Butler, J.L.; Pickett, J.B.

    1982-03-01

    A quantitative merit evaluation which assesses the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste forms is presented. The reference borosilicate glass process is rated as the simplest, followed by FUETAP concrete. The other processes evaluated in order of increasing complexity were: glass marbles in a lead matrix, high-silica glass, crystalline ceramic (Synroc-D and tailored ceramic), and coated ceramic particles. Cost appraisals are summarized for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities

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

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  2. Final conditioning of high-level liquid radioactive waste

    International Nuclear Information System (INIS)

    Krause, H.

    1981-01-01

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

  3. Porous glass with high silica content for nuclear waste storage : preparation, characterization and leaching

    International Nuclear Information System (INIS)

    Aegerter, M.A.; Santos, D.I. dos; Ventura, P.C.S.

    1984-01-01

    Aqueous solutions simulating radioactive nuclear wastes (like Savanah River Laboratory) were incorporated in porous glass matrix with high silica content prepared by decomposition of borosilicate glass like Na 2 O - B 2 O 3 - SiO 2 . After sintering, the samples were submitted, during 28 days, to standard leaching tests MCC1, MCC5 (Soxhlet) and stagnating. The total weight loss, ph, as well as the integral and differential leaching rates and the accumulated concentrations in the leach of Si, Na, B, Ca, Mn, Al, Fe and Ni. The results are compared with the results from reference borosilicate glass, made by fusion, ceramic, synroc, concrets, etc... (E.G.) [pt

  4. Safe disposal of high-level radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-10-01

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

  5. Long term stability of yttria-stabilized zirconia waste forms. Stability for secular change of partitioned TRU waste composition by disintegration

    International Nuclear Information System (INIS)

    Kuramoto, Ken-ichi; Banba, Tsunetaka; Mitamura, Hisayoshi; Sakai, Etsuro; Uno, Masayoshi; Kinoshita, H.; Yamanaka, Shinsuke

    1999-01-01

    In this study, the stability of YSZ waste forms for secular change of partitioned TRU waste composition by disintegration, one of important terms in long-term stability, is the special concern. Designed amount of waste and YSZ powder were mixed and sintered. These TRU waste forms were submitted to tests of phase stability, chemical durability, mechanical property and compactness. The results were compared with those of another YSZ waste forms, non-radioactive Ce and/or Nd doped YSZ samples, and glass and Synroc waste forms. Experimental results show following: (1) Phase stability of (Np+Am)-, (Np+U)-, and (Np+U+Bi)-doped YSZ waste forms could be maintained of that of the initial Np+Am-doped YSZ waste form permanently even when the composition of partitioned TRU waste were changed by disintegration. (2) Secular change also accelerated volume increase of YSZ waste forms as well as alpha-decay damage. (3) Hv, E and K IC of (Np+U)- and (Np+U+Bi)-doped YSZ waste forms were independent of the secular change of the partitioned TRU waste composition by disintegration. (4) Mechanical properties of YSZ waste forms were more than those of a glass and Synroc waste forms. (5) Compactness of YSZ waste forms was good as waste forms for the partitioned TRU wastes. (J.P.N.)

  6. Leach testing of waste forms: interrelationship of ISO and MCC type tests

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1982-01-01

    Leach testing experiments were conducted on SYNROC-D material to examine the parameters which affect leaching results and to measure the activation energy for leaching of elements from SYNROC-D. Measured leach rates were found to be controlled by precipitation of insoluble phases for those tests where the sample surface area to volume of leachant (SA/V) multiplied by leaching time (t) exceeded 0.3 cm -1 d for leach tests at 90 0 C. In these cases the apparent activation energy for leaching was approximately 10 kcal/mole based on Na and Si data. For leach tests at 90 0 C with (Sa/V)(t) less than 0.2 cm -1 d, the activation energy for Na and Si dissolution was 18.5 kcal/mole for sample S29 and 14.5 kcal/mole for sample LSO4. The effect of sample geometry was investigated by leaching a series of crushed samples of different grain size. The results support the view that geometric surface area should be used in leach rate calculations rather than gas adsorption BET surface area. Comparison of results on S29 leaching of crushed samples and monoliths show that data from MCC-1 and ISO type leach tests may be directly compared when the data are examined at constant (SA/V)(t). 5 figures, 13 tables

  7. Synthesis and characterization of CaTiO3 powder by combustion synthesis process

    International Nuclear Information System (INIS)

    Jung, C. W.; Shin, H. C.; Park, J. Y.; Lee, H. G.; Kim, H. Y.; Hong, K. W.

    2000-01-01

    Synroc is considered as a one of the most promising candidate for HLW solidification. CaTiO 3 , perovskite, which is a component of Synroc, can immobilize lanthanide and actinides by forming solid solutions. Generally most of the radioactive wastes elements were treated as a nitrate form. Therefore, the combustion process using metal nitrates as reactant materials can be easily applied to immobilize the radioactive waste elements. In this study, the feasibility of preparing fine, single-phase powders of multi-component oxide by a combustion process was investigated. Generally, the powder synthesized by combustion process showed different characteristics depending on the type and amount of fuel. And the spherical CaTiO 3 particles were directly prepared from the aqueous solution by an ultrasonic mist combustion process using an ultrasonic nebulizers as mist generators. The particles prepared with simple spray pyrolysis method using nitrate solution without fuel as precursor solution showed porous and hollow morphology, while the particles prepared with precursor solutions containing fuel showed dense solid morphology. Among various kinds of fuel tested, glycine showed the best result in reaction kinetics and crystalline phase purity

  8. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    McCoy, H.E.

    1983-01-01

    Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 300 0 C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 300 0 C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800 0 C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts

  9. Improving iron-enriched basalt with additions of ZrO2 and TiO2

    International Nuclear Information System (INIS)

    Reimann, G.A.; Kong, P.C.

    1993-06-01

    The iron-enriched basalt (IEB) waste form, developed at the Idaho National Engineering Laboratory a decade ago, was modified to IEB4 by adding sufficient ZrO 2 and TiO 2 to develop crystals of zirconolite upon cooling, in addition to the crystals that normally form in a cooling basalt. Zirconolite (CaZrTi 2 O 7 ) is an extremely leach-resistant mineral with a strong affinity for actinides. Zirconolite crystals containing uranium and thorium have been found that have endured more than 2 billion years of natural processes. On this basis, zirconolite was considered to be an ideal host crystal for the actinides contained in transuranic (TRU)-contaminated wastes. Crystals of zirconolite were developed in laboratory melts of IEB4 that contained 5% each of ZrO 2 and TiO 2 and that were slow-cooled in the 1200--1000 degrees C range. When actinide surrogates were added to IEB4, these oxides were incorporated into the crystals of zirconolite rather than precipitating in the residual glass phase. Zirconolite crystals developed in IEB4 should stabilize and immobilize the dilute TRUs in heterogeneous, buried low-level wastes as effectively as this same phase does in the various formulations of Synroc used for the more concentrated TRUs encountered in high-level wastes. Synroc requires hot-pressing equipment, while IEB4 precipitates zirconolite from a cooling basaltic melt

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  11. ANSTO. Annual Report 1993-1994

    International Nuclear Information System (INIS)

    1994-09-01

    Scientific highlights during 1993-1994 financial year at the Australian Nuclear Science and Technology Organization (ANSTO) as outlined in the Annual Report include: a new Synroc facility was commissioned to provide microsphere feedstocks and confirmed the choice of the dry precursor route for the Synroc conceptual plant; joint research in plasma immersion ion implantation (PI3) with the Technical University of Clausthal, Germany; the design and manufacture of a prototype ceramic knee prosthesis; international collaboration established in the use of accelerator techniques to measure aerosol pollution; the discovery of new low temperature phases of palladium deuteride which crystal structures were determined using neutron scattering; elucidate the controversial age of the Venafro Chessmen using Accelerator Mass Spectrometry. Achievements in the biomedical fields included: the successful clinical evaluation of 123 I-iododexetimide in patients with Alzheimer's disease or frontal lobe epilepsy and the completed clinical trial of technetium-99m 3B6/22 antibody for the diagnosis of lung cancer. ANSTO has also completed two studies on the treatment of contaminated wastes arising from the flooding of uranium mines in Germany and advised the German Ministry of Economics on treatment options, developed new processes for the production of high purity cerium compounds from monazite concentrates and a computer software to assess the likelihood of a pollution release from the failure of industrial equipment and containment or clean-up systems. Details are also given of the Corporate and Information Services activities. The financial statements for the year under review is included. ills., tabs

  12. Improving iron-enriched basalt with additions of ZrO{sub 2} and TiO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Reimann, G.A.; Kong, P.C.

    1993-06-01

    The iron-enriched basalt (IEB) waste form, developed at the Idaho National Engineering Laboratory a decade ago, was modified to IEB4 by adding sufficient ZrO{sub 2} and TiO{sub 2} to develop crystals of zirconolite upon cooling, in addition to the crystals that normally form in a cooling basalt. Zirconolite (CaZrTi{sub 2}O{sub 7}) is an extremely leach-resistant mineral with a strong affinity for actinides. Zirconolite crystals containing uranium and thorium have been found that have endured more than 2 billion years of natural processes. On this basis, zirconolite was considered to be an ideal host crystal for the actinides contained in transuranic (TRU)-contaminated wastes. Crystals of zirconolite were developed in laboratory melts of IEB4 that contained 5% each of ZrO{sub 2} and TiO{sub 2} and that were slow-cooled in the 1200--1000{degrees}C range. When actinide surrogates were added to IEB4, these oxides were incorporated into the crystals of zirconolite rather than precipitating in the residual glass phase. Zirconolite crystals developed in IEB4 should stabilize and immobilize the dilute TRUs in heterogeneous, buried low-level wastes as effectively as this same phase does in the various formulations of Synroc used for the more concentrated TRUs encountered in high-level wastes. Synroc requires hot-pressing equipment, while IEB4 precipitates zirconolite from a cooling basaltic melt.

  13. An assessment of methods for immobilizing reprocessed radioactive waste

    International Nuclear Information System (INIS)

    Murthy, M.K.; Baranyi, A.D.

    1980-05-01

    Nuclear waste forms presently used for the disposal of high-level wastes and other potential waste forms under development were studied using information available in the literature and by visits to the laboratories. The following waste forms were considered: Borosilicate glass, high-silica glass, glassceramics, supercalcine ceramics, synroc ceramics, borosilicate glass beads in a metal matrix, supercalcine and synroc ceramics in a metal matrix and coated ceramics. The following conclusions have been reached: To date the best developed wasteform, both in terms of overall product quality and process development, is monolithic borosilicate glass. However, hydrothermal instability is a major concern. Borosilicate glass in metal matrix waste form has better properties than monolithic borosilicate glass waste form. The process has been proven on a pilot scale. Hence, it is considered very close to monolithic glass in terms of overall development. The product qualities of the other waste forms are better than borosilicate glass. However, process development for these alternative waste forms is still in a conceptual stage. The technological basis for processing ceramic waste forms exists in a well developed state. Nevertheless, adaptation of the technology to continuous hot-cell operation, although feasible, has not been demonstrated. In view of the product potential of ceramic waste forms it is felt that their development should be given emphasis at this time. (auth)

  14. Accelerated damage studies of titanate ceramics containing simulated PW-4b and JW-A waste

    International Nuclear Information System (INIS)

    Hart, K.P.; Vance, E.R.; Lumpkin, G.R.; Mitamura, H.; Matsumoto, S.; Banba, T.

    1999-01-01

    Ceramic waste forms are affected by radiation damage, primarily arising from aloha-decay processes that can lead to volume expansion and amorphization of the component crystalline phases. The understanding of the extent and impact of these effects on the overall durability of the waste form is critical to the prediction of their long-term performance under repository conditions. Since 1985 ANSTO and JAERI have carried out joint studies on the use of 244 Cm to simulate alpha-radiation damage in ceramic waste forms. These studies have focussed on synroc formulations doped with simulated PW-4b and JW-A wastes. The studies have established the relationship between density change and irradiation levels for Synroc containing JW-A and PW-4b wastes. The storage of samples at 200 C halves the rate of decrease in the density of the samples compared to that measured at room temperature. This effect is consistent with that found for natural samples where the amorphization of natural samples stored under crustal conditions is lower, by factors between 2 and 4, than that measured for samples from accelerated doping experiments stored at room temperature. (J.P.N.)

  15. Immobilization in ceramic waste forms of the residues from treatment of mixed wastes

    International Nuclear Information System (INIS)

    Oversby, V.M.; van Konynenburg, R.A.; Glassley, W.E.; Curtis, P.G.

    1993-11-01

    The Environmental Restoration and Waste Management Applied Technology Program at LLNL is developing a Mixed Waste Management Facility to demonstrate treatment technologies that provide an alternative to incineration. As part of that program, we are developing final waste forms using ceramic processing methods for the immobilization of the treatment process residues. The ceramic phase assemblages are based on using Synroc D as a starting point and varying the phase assemblage to accommodate the differences in chemistry between the treatment process residues and the defense waste for which Synroc D was developed. Two basic formulations are used, one for low ash residues resulting from treatment of organic materials contaminated with RCRA metals, and one for high ash residues generated from the treatment of plastics and paper products. Treatment process residues are mixed with ceramic precursor materials, dried, calcined, formed into pellets at room temperature, and sintered at 1150 to 1200 degrees C to produce the final waste form. This paper discusses the chemical composition of the waste streams and waste forms, the phase assemblages that serve as hosts for inorganic waste elements, and the changes in waste form characteristics as a function of variation in process parameters

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  17. Composite quarterly technical report: long-term high-level waste technology, October-December 1980

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-04-01

    The technical information in this report summarizes work performed at participating sites to immobilize high-level radioactive wastes. The areas reported are in: program management and support; waste preparation; waste fixation; and final handling. Majority of the studies were in the area of waste fixation, some of which are: leaching tests of ceramic forms, high silica glass, graphite powder and other carbon preparations; viscosity measurements for a range of waste-glass compositions from references borosilicate glass to high-alumina glasses; neutron activation analysis for measuring leach rates; preparation of SYNROC D spheres; formulations for preparing ceramics from defense waste composition; development of a pilot-scale glass melter, and kinetic studies of slag formation in glass melters

  18. Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004. Part 1: Contributions by participants in the co-ordinated research project on chemical durability and performance assessment under simulated repository conditions

    International Nuclear Information System (INIS)

    2007-07-01

    This publication contains the results of an IAEA Coordinated Research Project (CRP). It provides a basis for understanding the potential interactions of waste form and repository environment, which is necessary for the development of the design and safety case for deep disposal. Types of high level waste matrices investigated include spent fuel, glasses and ceramics. Of particular interest are the experimental results pertaining to ceramic forms such as SYNROC. This publication also outlines important areas for future work, namely, standardized, collaborative experimental protocols for package-release studies, structured development and calibration of predictive models linking the performance of packaged waste and the repository environment, and studies of the long term behaviour of the wastes, including active waste samples. It comprises 15 contributions of the participants on the Coordinated Research Project which are indexed individually.

  19. The structures and stability of media intended for the immobilization of high level radioactive waste

    International Nuclear Information System (INIS)

    Tempest, P.A.

    1979-05-01

    High level radioactive waste contains about 40 different elements and, in time, many of these elements are transformed by radioactive decay into different-sized atoms with new chemical properties. The suitability of ordered crystal structures and unordered glass structures as media for immobilising the waste elements is compared. The structural properties of a mixture of synthetic minerals (SYNROC) are described and the various minerals' ability to accommodate ions of different radii and charge assessed. Similary the unordered structure of glass is examined and the probability of the glass remaining non-crystalline during manufacture and storage taken into account. Alternative glassification technologies in the form of the French AVM continuous process and the UK HARVEST batch processes are described and compared, and their likely effect on the structural properties of the final solid glass block considered. (author)

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

    International Nuclear Information System (INIS)

    1982-03-01

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

  1. Factors affecting the release of radioactivity to the biosphere during deep geologic disposal of radioactive solids through underground water

    International Nuclear Information System (INIS)

    Solomah, A.G.

    1984-01-01

    The chemical alteration formed by ground water on the solidified radioactive waste during deep geologic disposal represents the most likely mechanism by which dangerous radioactive species could be reintroduced into the biosphere. Knowing the geologic history of the repository, the chemistry of the ground water and the mechanisms involved in the corrosion of the radioactive solids can provide help to predict the long-term stability of these materials. The factors that must be considered in order to assess the safety and the risk associated with such a disposal strategy are presented. The leaching behavior of a solidified radioactive waste form called SYNROC-B (SYNthetic ROCks) is discussed. Different simulated ground water brines similar to those of the repository sites were prepared and used as the leaching media in leaching experiments

  2. Progress report on safety research of high-level waste management for the period April 1986 to March 1987

    International Nuclear Information System (INIS)

    Nakamura, Haruto; Tashiro, Shingo

    1987-08-01

    Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1986 are reviewed in the report. Topics in the three sections are as follows: 1) Non-radioactive research has been continued on Synroc irradiation and modellings of waste form leaching. 2) Research results are described in the section of Safety Evaluation for Geological Disposal on engineered barriers, field tests, safety assessment models, migration, natural analogue, seabed disposal and conceptual design of a repository. 3) Adsorption behaviour of plutonium on leach-containers and migration of leached cesium in a rock column are described in the section of Safety Examination of Vitrified Forms in the Hot Cells of WASTEF. (author)

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

    International Nuclear Information System (INIS)

    2007-10-01

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

  4. Nuclear energy and its future

    International Nuclear Information System (INIS)

    Cook, D.J.

    1990-01-01

    The status of nuclear power in the world and its future are briefly discussed. It is shown that nuclear power capacity is increasing in the Asian and Pacific rim region and that new reactor designs, with the increased emphasis on safety and standardisation, could make nuclear power a more acceptable option in the future. The author also outlines the Australian Nuclear Science and Technology Organization wide range of skills and facilities which are bringing the benefits of nuclear science and technology to Australia. These include: the development of Synroc as an advanced second generation waste management; production of radiotracers for biomedical researches and environmental problems; application of gamma irradiation in industry and of ion beam analysis in biology, archaeology, semi-conductor and environmental science. 2 tabs

  5. Nuclear power in perspective

    International Nuclear Information System (INIS)

    Ringwood, A.E.

    1980-01-01

    The nuclear power debate hinges upon three major issues: radioactive waste disposal, reactor safety and proliferation. An alternative strategy for waste disposal is advocated which involves disposing of the radwaste (immobilized in SYNROC, a titanate ceramic waste form) in deep (4 km) drill-holes widely dispersed throughout the entire country. It is demonstrated that this strategy possesses major technical (safety) advantages over centralized, mined repositories. The comparative risks associated with coal-fired power generation and with the nuclear fuel cycle have been evaluated by many scientists, who conclude that nuclear power is far less hazardous. Considerable improvements in reactor design and safety are readily attainable. The nuclear industry should be obliged to meet these higher standards. The most hopeful means of limiting proliferation lies in international agreements, possibly combined with international monitoring and control of key segments of the fuel cycle, such as reprocessing

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

    International Nuclear Information System (INIS)

    Banba, Tsunetaka; Kamizono, Hiroshi; Mitamura, Hisayoshi

    1992-02-01

    The recent studies of high-level radioactive waste form at Japan Atomic Energy Research Institute can be classified into the following three categories; (1) Study on the leaching behavior of the nuclear waste glass placing the focus on the alteration layer and the chemical composition of leachant for the prediction of the long-term corrosion of the waste glass. (2) Study on the radiation (alpha-radiation) effects which have relation to the long-term stability of the nuclear waste glass. (3) Study on the long-term self-irradiation damage of a SYNROC waste form using a curium-doped sample. In the present report, the recent results corresponding to the above categories are described. (author)

  7. Development of high-level waste solidification technology 1

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

  8. Progress report on safety research of high-level waste management for the period April 1987 to March 1988

    International Nuclear Information System (INIS)

    Nakamura, Haruto; Tashiro, Shingo

    1988-10-01

    Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1987 are reviewed in the three sections of the report. The topics are as follows: 1) On performance and durability of waste forms and engineered barrier materials, accelerated alpha radiation stability of glass form and Synroc has been investigated and stress corrosion cracking of canister materials was examined under simulated conditions. 2) Sorption of 237 Np on granite samples and behavior of iron during weathering of granites were studied with respect to safety evaluation for geological disposal. 3) Actual waste was transported from the Tokai Reprocessing Plant and hot operation using the actual waste was initiated at WASTEF. (author)

  9. Self-propagating synthesis and aqueous durability of Nd-bearing zirconolite-rich composites using Ca(NO{sub 3}){sub 2} as the oxidant

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Kuibao, E-mail: xiaobao320@163.com [State Key Laboratory Cultivation Base for Nonmetal Composite and Functional Materials, Southwest University of Science and Technology, Mianyang, Sichuan 621010 (China); National Defense Key Discipline Lab of Nuclear Waste and Environmental Safety, Southwest University of Science and Technology, Mianyang 621010 (China); He, Shihong [State Nuclear Power Research Institute, Beijing 100029 (China); Yin, Dan; Peng, Le; Wu, Jingjun [State Key Laboratory Cultivation Base for Nonmetal Composite and Functional Materials, Southwest University of Science and Technology, Mianyang, Sichuan 621010 (China)

    2016-09-15

    Synroc is recognized as the second-generation waste form for safety disposal of high-level radioactive waste (HLW). In this study, zirconolite-rich Synroc waste form was readily synthesized by self-propagating high-temperature plus quick pressing (SHS/QP) using Ca(NO{sub 3}){sub 2} as the oxidant and Ti as the reductant. As the surrogate of trivalent actinides, Nd{sub 2}O{sub 3} was introduced to equally substitute the Ca and Zr sites of zirconolite with nominal stoichiometry of Ca{sub 1−x}Zr{sub 1−x}Nd{sub 2x}Ti{sub 2}O{sub 7}. The results demonstrate that zirconolite, perovskite and pyrochlore (Ca{sub 2}Ti{sub 2}O{sub 6}) coexist as the ceramic components after SHS reaction. The introduction of Nd{sub 2}O{sub 3} promotes the formation of perovskite. Nd is mostly incorporated into the Ca sites of these phases. The normalized elemental leaching rates of Ca and Nd are fairly constant in low values of 1.80 × 10{sup −2} g m{sup −2} d{sup −1} and 6.12 × 10{sup −4} g m{sup −2} d{sup −1} after 42 days. - Highlights: • Zirconolite-rich composite was synthesized by SHS using Ca(NO{sub 3}){sub 2} as the oxidant. • Nd{sub 2}O{sub 3} was successfully immobilized into the crystal structure of this waste form. • Nd was mostly incorporated into the Ca sites of zirconolite, perovskite and pyrochlore. • The normalized leaching rates of Ca and Nd are in relatively low values.

  10. Selection of appropriate conditioning matrices for the safe disposal of radioactive waste

    International Nuclear Information System (INIS)

    Vance, E.R.

    2002-01-01

    The selection of appropriate solid conditioning matrices or wasteforms for the safe disposal of radioactive waste is dictated by many factors. The overriding issue is that the matrix incorporating the radionuclides, together with a set of engineered barriers in a near-surface or deep geological repository, should prevent significant groundwater transport of radionuclides to the biosphere. For high-level waste (HLW) from nuclear fuel reprocessing, the favored matrices are glasses, ceramics and glass-ceramics. Borosilicate glasses are presently being used in some countries, but there are strong scientific arguments why ceramics based on assemblages of natural minerals are advantageous for HLW. Much research has been carried out in the last 40 years around the world, and different matrices are more suitable than others for a given waste composition. However a major stumbling block for HLW immobilisation is the mall number of approved geological repositories for such matrices. The most appropriate matrices for Intermediate and low-level wastes are contentious and the selection criteria are not very well defined. The candidate matrices for these latter wastes are cements, bitumen, geopolymers, glasses, glass-ceramics and ceramics. After discussing the pros and cons of various candidate matrices for given kinds of radioactive wastes, the SYNROC research program at ANSTO will be briefly surveyed. Some of the potential applications of this work using a variety of SYNROC derivatives will be given. Finally the basic research program at ANSTO on radioactive waste immobilisation will be summarised. This comprises mainly work on solid state chemistry to understand ionic valences and co-ordinations for the chemical design of wasteforms, aqueous durability to study the pH and temperature dependence of solid-water reactions, radiation damage effects on structure and solid-water reactions. (Author)

  11. Preliminary evaluation of alternative waste form solidification processes. Volume I. Identification of the processes

    International Nuclear Information System (INIS)

    Treat, R.L.; Nesbitt, J.F.; Blair, H.T.; Carter, J.G.; Gorton, P.S.; Partain, W.L.; Timmerman, C.L.

    1980-04-01

    This document contains preconceptual design data on 11 processes for the solidification and isolation of nuclear high-level liquid wastes (HLLW). The processes are: in-can glass melting (ICGM) process, joule-heated glass melting (JHGM) process, glass-ceramic (GC) process, marbles-in-lead (MIL) matrix process, supercalcine pellets-in-metal (SCPIM) matrix process, pyrolytic-carbon coated pellets-in-metal (PCCPIM) matrix process, supercalcine hot-isostatic-pressing (SCHIP) process, SYNROC hot-isostatic-pressing (SYNROC HIP) process, titanate process, concrete process, and cermet process. For the purposes of this study, it was assumed that each of the solidification processes is capable of handling similar amounts of HLLW generated in a production-sized fuel reprocessing plant. It was also assumed that each of the processes would be enclosed in a shielded canyon or cells within a waste facility located at the fuel reprocessing plant. Finally, it was assumed that all of the processes would be subject to the same set of regulations, codes and standards. Each of the solidification processes converts waste into forms that may be acceptable for geological disposal. Each process begins with the receipt of HLLW from the fuel reprocessing plant. In this study, it was assumed that the original composition of the HLLW would be the same for each process. The process ends when the different waste forms are enclosed in canisters or containers that are acceptable for interim storage. Overviews of each of the 11 processes and the bases used for their identification are presented in the first part of this report. Each process, including its equipment and its requirements, is covered in more detail in Appendices A through K. Pertinent information on the current state of the art and the research and development required for the implementation of each process are also noted in the appendices

  12. ANSTO. Annual Report 1993-1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-09-01

    Scientific highlights during 1993-1994 financial year at the Australian Nuclear Science and Technology Organization (ANSTO) as outlined in the Annual Report include: a new Synroc facility was commissioned to provide microsphere feedstocks and confirmed the choice of the dry precursor route for the Synroc conceptual plant; joint research in plasma immersion ion implantation (PI3) with the Technical University of Clausthal, Germany; the design and manufacture of a prototype ceramic knee prosthesis; international collaboration established in the use of accelerator techniques to measure aerosol pollution; the discovery of new low temperature phases of palladium deuteride which crystal structures were determined using neutron scattering; elucidate the controversial age of the Venafro Chessmen using Accelerator Mass Spectrometry. Achievements in the biomedical fields included: the successful clinical evaluation of {sup 123}I-iododexetimide in patients with Alzheimer`s disease or frontal lobe epilepsy and the completed clinical trial of technetium-99m 3B6/22 antibody for the diagnosis of lung cancer. ANSTO has also completed two studies on the treatment of contaminated wastes arising from the flooding of uranium mines in Germany and advised the German Ministry of Economics on treatment options, developed new processes for the production of high purity cerium compounds from monazite concentrates and a computer software to assess the likelihood of a pollution release from the failure of industrial equipment and containment or clean-up systems. Details are also given of the Corporate and Information Services activities. The financial statements for the year under review is included. ills., tabs.

  13. Performance of high level waste forms and engineered barriers under repository conditions

    International Nuclear Information System (INIS)

    1991-02-01

    The IAEA initiated in 1977 a co-ordinated research programme on the ''Evaluation of Solidified High-Level Waste Forms'' which was terminated in 1983. As there was a continuing need for international collaboration in research on solidified high-level waste form and spent fuel, the IAEA initiated a new programme in 1984. The new programme, besides including spent fuel and SYNROC, also placed greater emphasis on the effect of the engineered barriers of future repositories on the properties of the waste form. These engineered barriers included containers, overpacks, buffer and backfill materials etc. as components of the ''near-field'' of the repository. The Co-ordinated Research Programme on the Performance of High-Level Waste Forms and Engineered Barriers Under Repository Conditions had the objectives of promoting the exchange of information on the experience gained by different Member States in experimental performance data and technical model evaluation of solidified high level waste forms, components of the waste package and the complete waste management system under conditions relevant to final repository disposal. The programme includes studies on both irradiated spent fuel and glass and ceramic forms as the final solidified waste forms. The following topics were discussed: Leaching of vitrified high-level wastes, modelling of glass behaviour in clay, salt and granite repositories, environmental impacts of radionuclide release, synroc use for high--level waste solidification, leachate-rock interactions, spent fuel disposal in deep geologic repositories and radionuclide release mechanisms from various fuel types, radiolysis and selective leaching correlated with matrix alteration. Refs, figs and tabs

  14. Support for DOE program in mineral waste-form development

    International Nuclear Information System (INIS)

    Palmour, H. III; Hare, T.M.; Russ, J.C.; Batchelor, A.D.; Paisley, M.J.; Freed, L.E.

    1982-09-01

    This research investigation relates to sintered simulation ceramic waste forms of the generic SYNROC compositional type. Though they have been formulated with simulated wastes only, they serve as prototypes for potential hot, processed, crystalline waste forms whose combined thermodynamic stability and physical integrity are considered to render them capable of long-term imobilization of high-level radwastes under deep geologic disposal conditions. The problems involved are nontrivial, largely because of the very complex nature of the radwastes: a typical waste stream would contain more than 31 cation species. When the stabilizing matrix constituents are included, the final batch composition must successfully account (and find substitutional homes for some 35 different cation species. One of the important objectives of this study thus has been to develop a computer-based method for simulating these complex ion substitutions, and for calculating the resultant phase demands and batch formulations. Primary goals of the study have been (1) use of that computer simulation capability to incorporate rationally the radwaste ions from a specific waste stream (PW-7a) into the available SYNROC lattice sites and (2) utilization of existing ceramic processing and sintering methodologies to assure (and to understand) the attainment of high density, fine microstructure, full phase development and other features of the sintered product which are known to relate directly to its integrity and leach resistance. Though improved resistance to leaching has been a continuing goal, time and budget constraints have precluded initiation of any leachability studies of these new compositions during this contract period. 27 references, 15 figures, 6 tables

  15. Charge compensation and the incorporation of cerium in zirconolite and perovskite

    International Nuclear Information System (INIS)

    Begg, B.D.; Vance, E.R.; Lumpkin, G.R.

    1998-01-01

    Full text: Synroc is a mineral-analogue based titanate ceramic, consisting of a series of extremely stable, mutually compatible phases capable of incorporating HLW elements within their crystal structures. Waste elements are incorporated into the each of the Synroc phases via a substitutional solid solution mechanism. A given waste element is substituted directly for a host matrix element, of a similar ionic size, and where a charge imbalance exists between the waste and the host ions, suitable charge compensation is made to maintain overall charge neutrality. Charge compensation may take the form of an additional ion of appropriate charge substituting on either the same or a separate site, in such a manner so as to offset the original charge imbalance. In this way, waste ions are chemically bonded into the crystal structure of the durable host Synroc phase. The major rare earth/actinide-bearing Synroc phase is zirconolite. Previously we have reported on the incorporation of both cerium, which was used as a non-radioactive simulant for plutonium, and plutonium in zirconolite. We demonstrated how the valence of both ions can be varied by changing the firing atmosphere without significantly altering the composition of the zirconolite. This raised a number of significant questions about the nature of charge compensation at work in these zirconolites. In an effort to further investigate the charge compensation mechanisms at work in these cerium- and plutonium-doped zirconolites, it was decided to examine the incorporation of Ce in the simpler, but closely related, perovskite (CaTiO 3 ) system in addition to making further studies of Ce-doped zirconolites. Of course perovskite is also a component of Synroc which is also capable of incorporating significant amounts of rare earths and actinides. In an analogous way to the zirconolite series, the Ce was incorporated on the Ca site, with specific Ce valence states being targeted via the provision of appropriate amounts of

  16. 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

  17. Final report on initial samples supplied by LLNL for task 3.3 binder burnout and sintering schedule optimisation

    Energy Technology Data Exchange (ETDEWEB)

    Walls, P

    1999-01-04

    Sixteen of the twenty-one samples have been investigated using the scanning laser dilatometer. This includes all three types of samples with different preparation routes and organic content. Cracks were observed in all samples, even those only heated to 300 C. It was concluded that the cracking was occurring in the early part of the heat treatment before the samples reached 300 C. Increase in the rate of dilation of the samples occurred above 170 C which coincided with the decomposition of the binder/wax additives as determined by differential thermal analysis. A comparison was made with SYNROC C material (Powder Run 143), samples of which had been CIPed and green machined to a similar diameter and thickness as the 089 mm SRTC pucks. These samples contained neither binder nor other organic processing aids and had been kept in the same desiccator as the SRTC samples. The CIPed Synroc C samples sintered to high density with zero cracks. As the cracks made up only a small contribution to the change in diameter of the sample compared to the sintering shrinkage, useful information could still be gained from the runs. The sintering curves showed that there was much greater shrinkage of the Type III samples containing only the 5% PEG binder compared to the Type I which contained polyolefin wax as processing aid. Slight changes in gradient of the sintering curve were observed, however, due to the masking effect of the cracking, full analysis of the sintering kinetics cannot be conducted. Even heating the samples to 300 C at 1.0 or 0.5 C/min could not prevent crack formation. This indicated that heating rate was not the critical parameter causing cracking of the samples. Sectioning of green bodies revealed the inhomogeneous nature of the binder/lubricant distribution in the samples. Increased homogeneity would reduce the amount of binder/lubricant required, which should in turn, reduce the degree of cracking observed during heating to the binder burnout temperature. A

  18. Final Report on Initial Samples Supplied by LLNL for Task 3.3 Binder Burnout and Sintering Schedule Optimisation

    Energy Technology Data Exchange (ETDEWEB)

    Walls, P

    1999-01-04

    Sixteen of the twenty-one samples have been investigated using the scanning laser dilatometer. This includes all three types of samples with different preparation routes and organic content. Cracks were observed in all samples, even those only heated to 300 C. It was concluded that the cracking was occurring in the early part of the heat treatment before the samples reached 300 C. Increase in the rate of dilation of the samples occurred above 170 C which coincided with the decomposition of the binder/wax additives as determined by differential thermal analysis. A comparison was made with SYNROC C material (Powder Run 143), samples of which had been CIPed and green machined to a similar diameter and thickness as the 089mm SRTC pucks. These samples contained neither binder nor other organic processing aids and had been kept in the same desiccator as the SRTC samples. The CIPed Synroc C samples sintered to high density with zero cracks. As the cracks made up only a small contribution to the change in diameter of the sample compared to the sintering shrinkage, useful information could still be gained from the runs. The sintering curves showed that there was much greater shrinkage of the Type III samples containing only the 5% PEG binder compared to the Type I which contained polyolefin wax as processing aid. Slight changes in gradient of the sintering curve were observed, however, due to the masking effect of the cracking, full analysis of the sintering kinetics cannot be conducted. Even heating the samples to 300 C at 1.0 or 0.5 C/min could not prevent crack formation. This indicated that heating rate was not the critical parameter causing cracking of the samples. Sectioning of green bodies revealed the inhomogeneous nature of the binder/lubricant distribution in the samples. Increased homogeneity would reduce the amount of binder/lubricant required, which should in turn, reduce the degree of cracking observed during heating to the binder burnout temperature. A

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

    International Nuclear Information System (INIS)

    Grover, J.R.

    1980-01-01

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

  20. Sol-gel technology applied to crystalline ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    Angelini, P.; Bond, W.D.; Caputo, A.J.; Mack, J.E.; Lackey, W.J.; Lee, D.A.; Stinton, D.P.

    1980-01-01

    The sol-gel process is being developed for the solidification and isolation of high-level nuclear fuel waste. Three gelation methods are being developed for producing alternative waste forms. These include internal gelation for producing spheres of up to 1 mm diam suitable for coating, external gelation, and water extraction methods for producing material suitable for alternate ceramic processing. In this study internal gelation has been used to produce ceramic spheres of various alternative nuclear waste compositions. A gelation system capable of producing 100-g batches has been assembled and used for development. Waste forms containing up to 70 wt % simulated Savannah River Plant waste have been produced. Dopants such as Cs, Sr, Nd, Ru, and Mo were used in some experiments to observe side waste streams and sintering effects. Synroc microspheres were coated with both low-density carbon, high-density impermeable carbon, high-temperature dense SiC, and SiC deposited at temperatures near 900 0 C. Other gelation methods and other alternative waste forms are being developed

  1. Development of thermal conditioning technology for alpha-contaminated wastes

    International Nuclear Information System (INIS)

    Kim, Joon Hyung; Kim, H. Y.; Kim, J. G.

    2001-04-01

    To develop a thermal conditioning technology for alpha-contaminated wastes, which are presumed to generate from pyrochemical processing of spent fuel, research on the three different fields have been performed; incineration, off-gas treatment, and vitrification/cementation technology. Through the assessment on the amount of alpha-contaminated waste and incineration characterises, an oxygen-enriched incineration process, which can greatly reduce the off-gas volume, was developed by our own technology. Trial burn test with paper waste resulted in a reduction of off-gas volume by 3.5. A study on the behavior and adsorption of nuclides/heavy metals at high-temperature was performed to develop an efficient removal technology. Off-gas treatment technologies for radioiodine at high-temperature and 14 CO 2 , acidic gases, and radioactive gaseous wastes such as Xe/Kr at room temperature were established. As a part of development of high-level waste solidification technology, manufacture of high-frequency induction melter, fabrication and characterization of base-glass media fabricated with spent HEPA filter medium, and development of titanate ceramic material as a precursor of SYNROC by a self-combustion method were performed. To develop alpha-contaminated waste solidification technology, a process to convert periodontal in the cement matrix to calcite with SuperCritical Carbon Dioxide (SCCD) was manufactured. The SCCD treatment enhanced the physicochemical properties of cement matrices, which increase the long-term integrity of cement waste forms during transportation and storage

  2. Preliminary assessment of nine waste-form products/processes for immobilizing transuranic wastes

    International Nuclear Information System (INIS)

    Crisler, L.R.

    1980-09-01

    Nine waste-form processes for reduction of the present and projected Transuranic (TRU) waste inventory to an immobilized product have been evaluated. Product formulations, selected properties, preparation methods, technology status, problem areas needing resolution and location of current research development being pursued in the United States are discussed for each process. No definitive utility ranking is attempted due to the early stage of product/process development for TRU waste containing products and the uncertainties in the state of current knowledge of TRU waste feed compositional and quantitative makeup. Of the nine waste form products/processes included in this discussion, bitumen and cements (encapsulation agents) demonstrate the degree of flexibility necessary to immobilize the wide composition range present in the TRU waste inventory. A demonstrated process called Slagging Pyrolysis Incineration converts a varied compositional feed (municipal wastes) to a ''basalt'' like product. This process/product appears to have potential for TRU waste immobilization. The remaining waste forms (borosilicate glass, high-silica glass, glass ceramics, ''SYNROC B'' and cermets) have potential for immobilizing a smaller fraction of the TRU waste inventory than the above discussed waste forms

  3. Annual report on operation, utilization and technical development of hot laboratories. From April 1, 1996 to March 31, 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    This report describes activities, in fiscal year 1996, of the Reactor Fuel Examination Facility (RFEF), the Research Hot Laboratory (RHL) and the Waste Safety Testing Facility (WASTEF) which belong to the Department of Hot laboratories. In the RFEF, Post-Irradiation Examinations (PIEs) of PWR fuel assemblies irradiated in the Takahama Unit 3, a BWR fuel assembly irradiated in the Fukusima Daini Unit have been performed. Also, PIEs of assembly materials irradiated in the Fugen Reactor have been carried out. To support R and D works in JAERI, refabrication of segmented fuel rods have been done using irradiated LWR fuel rods for pulse irradiation in the NSRR and re-irradiation tests in the JMTR. PIEs have been performed on high burnup fuel rods and ROX fuel rods. For the RHL, PIEs have been performed on segment fuels irradiated in the NSRR, fuels and materials for HTTR, standard fuels for JRR-3M and materials for nuclear fusion reactor. In addition, a monitoring test of fuel elements in accordance with the surveillance program of the Magnox reactor of the Japan Atomic Power Corporation has been continued. In the WASTEF, leaching tests on TRU in simulated glass forms and a low flow rate tests on glass waste forms have been carried out. The examinations of alpha damage acceleration for the Synroc waste forms have also been performed. (author)

  4. Synthesis, characterization and structural refinement of polycrystalline uranium substituted zirconolite

    International Nuclear Information System (INIS)

    Shrivastava, O.P.; Narendra Kumar; Sharma, I.B.

    2005-01-01

    Ceramic precursors of Zirconolite (CaZrTi 2 O 7 ) family have a remarkable property of substitution Zr 4+ cationic sites. This makes them potential material for nuclear waste management in 'synroc' technology. In order to simulate the mechanism of partial substitution of zirconium by tetravalent actinides, a solid phase of composition CaZr 0.95 U 0.5 Ti 2 O 7 has been synthesized through ceramic route by taking calculated quantities of oxides of Ca, Ti and nitrates of uranium and zirconium respectively. Solid state synthesis has been carried out by repeated pelletizing and sintering the finely powdered oxide mixture in a muffle furnace at 1050 degC. The polycrystalline solid phase has been characterized by its typical powder diffraction pattern. Step analysis data has been used for ab initio calculation of structural parameters. The uranium substituted zirconolite crystallizes in monoclinic symmetry with space group C2/c (15). The following unit cell parameters have been calculated: a =12.4883(15), b =7.2448(5), c 11.3973(10) and β = 100.615(9)0. The structure was refined to satisfactory completion. The Rp and Rwp are found to be 7.48% and 9.74% respectively. (author)

  5. Status of plutonium ceramic immobilization processes and immobilization forms

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

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

  6. 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

  7. Japan-Australia co-operative program on research and development of technology for the management of high level radioactive wastes. Final report 1985 to 1998

    Energy Technology Data Exchange (ETDEWEB)

    Hart, K.; Vance, E.; Lumpkin, G. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia); Mitamura, H.; Banba, T. [Japan Atomic Energy Research Inst. Tokai, Ibaraki (Japan)

    1998-12-01

    The overall aim of the Co-operative Program has been to promote the exchange of information on technology for the management of High-Level Wastes (HLW) and to encourage research and development relevant to such technology. During the 13 years that the Program has been carried out, HLW management strategies have matured and developed internationally, and Japan has commenced construction of a domestic reprocessing and vitrification facility for HLW. The HLW management strategy preferred is a national decision. Many countries are using vitrification, direct disposal of spent fuel or a combination of both to handle their existing wastes whereas others have deferred the decision. The work carried out in the Co-operative Program provides strong scientific evidence that the durability of ceramic waste forms is not significantly affected by radiation damage and that high loadings of actinide elements can be incorporated into specially designed ceramic waste forms. Moreover, natural minerals have been shown to remain as closed systems for U and Th for up to 2.5 b y. All of these results give confidence in the ability of second generation waste forms, such as Synroc, to handle future waste arisings that may not be suitable for vitrification 87 refs., 15 tabs., 22 figs.

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-12-01

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

  9. Calcium titanium silicate based glass-ceramic for nuclear waste immobilisation

    Science.gov (United States)

    Sharma, K.; Srivastav, A. P.; Goswami, M.; Krishnan, Madangopal

    2018-04-01

    Titanate based ceramics (synroc) have been studied for immobilisation of nuclear wastes due to their high radiation and thermal stability. The aim of this study is to synthesis glass-ceramic with stable phases from alumino silicate glass composition and study the loading behavior of actinides in glass-ceramics. The effects of CaO and TiO2 addition on phase evolution and structural properties of alumino silicate based glasses with nominal composition x(10CaO-9TiO2)-y(10Na2O-5 Al2O3-56SiO2-10B2O3); where z = x/y = 1.4-1.8 are reported. The glasses are prepared by melt-quench technique and characterized for thermal and structural properties using DTA and Raman Spectroscopy. Glass transition and peak crystallization temperatures decrease with increase of CaO and TiO2 content, which implies the weakening of glass network and increased tendency of glasses towards crystallization. Sphene (CaTiSiO5) and perovskite (CaTiO3) crystalline phases are confirmed from XRD which are well known stable phase for conditioning of actinides. The microsturcture and elemental analysis indicate the presence of actinide in stable crystalline phases.

  10. A review and discussion of candidate ceramics for immobilization of high-level fuel reprocessing wastes

    International Nuclear Information System (INIS)

    Hayward, P.J.

    1982-08-01

    This review discusses and attempts to evaluate 11 of the leading ceramic processes for hosting the high-level and high-level plus medium-level wastes which would arise from the reprocessing of used UO 2 , (Th,Pu)O 2 and (Th,U)O 2 fuels. The wasteform materials considered include glass ceramics, supercalcine ceramics, SYNROC ceramics, 'stuffed glass', titanate ceramics, cermets, clay ceramics, cement-based materials and multibarrier wasteforms. Although no attempt has been made to rank these candidates in order of superiority, the conclusion is drawn that, of the materials proposed so far, a glass ceramic appears to be best suited to the Canadian program, taking into account durability in the potential environment of a flooded vault, ability to withstand radiation and transmutation damage without serious loss of durability, ability to accommodate variable waste compositions, and ease of processing and quality control. This conclusion does not necessarily apply to other national waste management programs. However, many of the points raised might be included in any critical assessment of alternative wasteform materials

  11. Australia's role in the nuclear fuel cycle. A report to the Prime Minister by the Australian Science and Technology Council (ASTEC)

    International Nuclear Information System (INIS)

    1984-05-01

    Results of an inquiry which was initiatd by the Australian Government in Novembr 1983 and which examined Australia's nuclear safeguards arrangements, the opportunities for Australia to advance the cause of nuclear non-proliferation, the adequacy of existing technology for the handling and disposal of radioactive wastes and ways in which Australia can further contribute to the development of safe disposal methods are presented. The report is also known as the Slatyer Inquiry. The 25 recommendations cover: export of Australia's uranium; participation in disarmament and arms control negotiations; the non-provision of nuclear items to non-NPT states; proposals for nuclear weapons free zones; guidelines for the supply of nuclear items; physical protection of nuclear material; regulating the storage and use of sensitive nuclear material; minimising the numbers of facilities such as enrichment and reprocessing plants; Australian participation in the nuclear fuel cycle; supporting safeguards operations by providing resources to the IAEA; supporting the IAEA's Program of Technical Assistance and Co-operation; participation in the IAEA; implementation of safeguards agreements; physical protection of nuclear materials during shipment; publicising administrative arrangements of safeguards agreements; limitation of releases of radioactive effluents; disposal of low and intermediate level wastes; standards for radiation exposure associated with uranium mining and milling; safety and environmental monitoring aspects of uranium mining and milling; a registry of radioactive tailings and waste disposal sites; ocean dumping; research into HLW disposal; support for R and D on Synroc and guidelines for HLW disposal

  12. Man, environment and nuclear energy

    International Nuclear Information System (INIS)

    Gardan, Jacques.

    1978-10-01

    The acceptability of nuclear fission as energy source is governed by three factors, economic, ecological and sociological. It is necessary to account first for the economic context and for the state of natural resources: gradual exhaustion of fossil fuels as a result of ever-increasing demands. The biological risk concept which determines the acceptable industrial application level is the second factor to be considered. The danger of radioactive contamination is almost inexistent except in the accident hypothesis, and power stations are built with excessive safeguards against hypothetical accidents. The idea of systematic processing of all working effluent to reduce radioactive waste discharge by several orders of magnitude (zero release principle) is being examined. At present, the waste discharge levels are always well below the limits set by the CIPR and present no danger to the population. The only serious problems seem to be the disposal of radioactive wastes and the plutonium non-proliferation question bound up with breeder reactors. Whereas vitrification, the new 'Synroc' process, offer some solution to the radioactive waste conditioning problem, responsibility for the proliferation of nuclear weapons rests with the human conscience alone. The development of nuclear power stations over several decades seems to present no inacceptable danger and offers the best compromise between growth and minimum risk requirements. The third factor to be accounted for is the opposition displayed by a fraction of the population to the development of nuclear energy for peaceful applications [fr

  13. Silica based gel as a potential waste form for high level waste from fuel reprocessing

    International Nuclear Information System (INIS)

    Ford, C.E.; Dempster, T.J.; Melling, P.J.

    1983-10-01

    To assess the feasibility of safe disposal of high-level radioactive waste as synthetic clay, or material that would react with ground water to form clay, experiments have been carried out to determine the hydrothermal crystallisation and leaching behaviour of silica based gels fired at 900 deg C. Crystallisation rates at a pressure of 500 bars and at temperatures below 400 deg C are negligible and this more or less precludes pre-disposal production of synthetic clay on the scale required. Leaching experiments suggest that the leach rates of Cs from gels by distilled water are higher than those of boro-silicate glasses and SYNROC at the lower temperatures that would be preferred for geological storage. However, amounts of bulk dissolution of gels may be lower than those of boro-silicate glasses. The initial leaching behaviour of gels might be considerably improved by hot compaction at 900 to 1000 deg C. Consideration of likely waste form dissolution behaviour in a repository environment suggests that gels of appropriate composition might perform as well as, or better than, boro-silicate glasses. A novel hypothetical plant is described that could produce the gel waste form on the scale required on a more or less continuous basis. (author)

  14. Titanate ceramics for immobilisation of uranium-rich radioactive wastes arising from {sup 99}Mo production

    Energy Technology Data Exchange (ETDEWEB)

    Carter, M.L.; Li, H. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia); Zhang, Y. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia)], E-mail: yzx@ansto.gov.au; Vance, E.R.; Mitchell, D.R.G. [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, PMB 1, Menai, Sydney, NSW 2232 (Australia)

    2009-02-28

    Uranium-rich liquid wastes arising from UO{sub 2} targets which have been neutron-irradiated to generate medical radioisotopes such as {sup 99m}Tc require immobilisation. A pyrochlore-rich hot isostatically pressed titanate ceramic can accommodate at least 40 wt% of such waste expressed on an oxide basis. In this paper, the baseline waste form composition (containing 40 wt% UO{sub 2}) was adjusted in two ways: (a) varying the UO{sub 2} loading with constant precursor oxide materials, (b) varying the precursor composition with constant waste loading of UO{sub 2}. This resulted in the samples having a similar phase assemblage but the amounts of each phase varied. The oxidation states of U in selected samples were determined using diffuse reflection spectroscopy (DRS) and electron energy loss spectroscopy (EELS). Leaching studies showed that there was no significant difference in the normalised elemental release rates and the normalised release rates are comparable with those from synroc-C. This demonstrates that waste forms based on titanate ceramics are robust and flexible for the immobilisation of U-rich waste streams from radioisotope processing.

  15. Electrical-conductivity measurements of leachates for the rapid assessment of wasteform corrosion resistance

    International Nuclear Information System (INIS)

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

    1982-01-01

    Measurements of the electrical conductivity of leachate solutions as a function of time can be used as an efficient, informative means of evaluation and comparison in the development of nuclear waste forms and in the preliminary analysis of their corrosion resistance in distilled water. Three separate applications of this technique are described in this work. These are: (1) its use in the optimization of the corrosion resistance of a crystalline wasteform (monazite); (2) a study of the protective ability of the surface layer (gel layer) which forms on the nuclear waste glass Frit 21 + 20 wt % SRW in distilled water; and (3) making comparisons of the overall corrosion resistance of three different nuclear wasteforms (i.e., monazite, SYNROC, and borosilicate glass). A complete solution analysis of the borosilicate glass leachate and a straightforward analysis of the conductivity results agree to within +-20%. In the absence of a complete, time consuming solution analysis, conductivity measurements can be used to estimate reliably the total ionic concentration in the leachate to within a factor of 2

  16. Status of plutonium ceramic immobilization processes and immobilization forms

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. Nuclear waste form risk assessment for US defense waste at Savannah River Plant. Annual report fiscal year 1980

    International Nuclear Information System (INIS)

    Cheung, H.; Jackson, D.D.; Revelli, M.A.

    1981-07-01

    Waste form dissolution studies and preliminary performance analyses were carried out to contribute a part of the data needed for the selection of a waste form for the disposal of Savannah River Plant defense waste in a deep geologic repository. The first portion of this work provides descriptions of the chemical interactions between the waste form and the geologic environment. We reviewed critically the dissolution/leaching data for borosilicate glass and SYNROC. Both chemical kinetic and thermodynamic models were developed to describe the dissolution process of these candidate waste forms so as to establish a fundamental basis for interpretation of experimental data and to provide directions for future experiments. The complementary second portion of this work is an assessment of the impacts of alternate waste forms upon the consequences of disposal in various proposed geological media. Employing systems analysis methodology, we began to evaluate the performance of a generic waste form for the case of a high risk scenario for a bedded salt repository. Results of sensitivity analysis, uncertainty analyses, and sensitivity to uncertainty analysis are presented

  18. Development of the plutonium oxide vitrification system

    International Nuclear Information System (INIS)

    Marshall, K.M.; Marra, J.C.; Coughlin, J.T.; Calloway, T.B.; Schumacher, R.F.; Zamecnik, J.R.; Pareizs, J.M.

    1998-01-01

    Repository disposal of plutonium in a suitable, immobilized form is being considered as one option for the disposition of surplus weapons-usable plutonium. Accelerated development efforts were completed in 1997 on two potential immobilization forms to facilitate downselection to one form for continued development. The two forms studied were a crystalline ceramic based on Synroc technology and a lanthanide borosilicate (LaBS) glass. As part of the glass development program, melter design activities and component testing were completed to demonstrate the feasibility of using glass as an immobilization medium. A prototypical melter was designed and built in 1997. The melter vessel and drain tube were constructed of a Pt/Rh alloy. Separate induction systems were used to heat the vessel and drain tube. A Pt/Rh stirrer was incorporated into the design to facilitate homogenization of the melt. Integrated powder feeding and off-gas systems completed the overall design. Concurrent with the design efforts, testing was conducted using a plutonium surrogate LaBS composition in an existing (near-scale) melter to demonstrate the feasibility of processing the LaBS glass on a production scale. Additionally, the drain tube configuration was successfully tested using a plutonium surrogate LaBS glass

  19. pH dependence of the aqueous dissolution rates of perovskite and zirconolite at 90 C

    International Nuclear Information System (INIS)

    McGlinn, P.J.; Hart, K.P.; Loi, E.H.; Vance, E.R.

    1995-01-01

    Perovskite and zirconolite are two of the major phases of the Synroc titanate mineral assemblage. Their aqueous durability under a range of pH conditions at 90 C has been examined. Solution analysis, electron microscopy and X-ray diffraction have been used to investigate the dissolution behavior of these phases, and a perovskite phase doped with Nd, Sr and Al, using buffered solutions at pH levels of 2.1, 3.7, 6.1, 7.9 and 12.9. After 43 days of leaching, Ca and Ti extractions from perovskite and zirconolite show only a weak pH-dependence. SEM investigation of the samples leached at pH 2.1, 6.1 and 12.9 showed that a titanaceous surface layer formed on the perovskite specimens. XRD analysis of the perovskite samples showed that anatase formed on the leached surface at acidic and neutral pHs, but not under alkaline conditions, and that minor amounts of rutile also formed. In the leached perovskite specimens doped with Nd, Sr and Al, no rule was found by XRD and anatase was only detected in the sample leached at pH 2.1. There were no detectable changes in the leached zirconolite samples examined by SEM and XRD

  20. Ceramic Hosts for Fission Products Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Peter C Kong

    2010-07-01

    Natural spinel, perovskite and zirconolite rank among the most leach resistant of mineral forms. They also have a strong affinity for a large number of other elements and including actinides. Specimens of natural perovskite and zirconolite were radioisotope dated and found to have survived at least 2 billion years of natural process while still remain their loading of uranium and thorium . Developers of the Synroc waste form recognized and exploited the capability of these minerals to securely immobilize TRU elements in high-level waste . However, the Synroc process requires a relatively uniform input and hot pressing equipment to produce the waste form. It is desirable to develop alternative approaches to fabricate these durable waste forms to immobilize the radioactive elements. One approach is using a high temperature process to synthesize these mineral host phases to incorporate the fission products in their crystalline structures. These mineral assemblages with immobilized fission products are then isolated in a durable high temperature glass for periods measured on a geologic time scale. This is a long term research concept and will begin with the laboratory synthesis of the pure spinel (MgAl2O4), perovskite (CaTiO3) and zirconolite (CaZrTi2O7) from their constituent oxides. High temperature furnace and/or thermal plasma will be used for the synthesis of these ceramic host phases. Nonradioactive strontium oxide will be doped into these ceramic phases to investigate the development of substitutional phases such as Mg1-xSrxAl2O4, Ca1-xSrxTiO3 and Ca1-xSrxZrTi2O7. X-ray diffraction will be used to establish the crystalline structures of the pure ceramic hosts and the substitution phases. Scanning electron microscopy and energy dispersive X-ray analysis (SEM-EDX) will be performed for product morphology and fission product surrogates distribution in the crystalline hosts. The range of strontium doping is planned to reach the full substitution of the divalent

  1. 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.

  2. ''Cs-tetra-ferri-annite:'' High-pressure and high-temperature behavior of a potential nuclear waste disposal phase

    International Nuclear Information System (INIS)

    Comodi, P.; Zanazzi, P.F.

    1999-01-01

    Structure deformations induced by pressure and temperature in synthetic Cs-tetra-ferri-annite 1M [Cs 1.78 (Fe 2+ 5.93 Fe 3+ 0.07 )(Si 6.15 Fe 3+ 1.80 Al 0.05 )O 20 (OH) 4 ], space group C2/m, were analyzed to investigate the capability of the mica structure to store the radiogenic isotopes 135 Cs and 137 Cs. Cs-tetra-ferri-annite is not a mineral name, but for the sake of brevity is used here to designate a synthetic analog of the mineral tetra-ferri-annite. The bulk modulus and its pressure derivative determined by fitting the unit-cell volumes between 0 a/nd 47 kbar to a third-order Birch-Murnaghan equation of state are K 0 = 257(8) kbar and K' 0 = 21(1), respectively. Between 23 C and 582 C, the a and b lattice parameters remain essentially unchanged, but the thermal expansion coefficient of the c axis is α c = 3.12(9) x 10 -5 degree C -1 . High pressure (P) and high temperature (T) produce limited internal strain in the structure. The tetrahedral rotation angle, α, is very small and does not change significantly throughout the P and T range investigated. Above 450 C in air, Cs-tetra-ferri-annite underwent an oxidation of octahedral iron in the M2cis site, balanced by the loss of H and shown by a decrease of the unit-cell volume. Independent isobaric data on thermal expansion and isothermal compressibility data define the geometric equation of state for Cs-tetra-ferri-annite. On the whole, the data confirm that the structure of Cs-tetra-ferri-annite may be a suitable candidate for the storage of large ions, such as Cs in the interlayer and should be considered as a potential Synroc component

  3. Perovskite as a matrix for incorporation of long-lived radionuclides

    International Nuclear Information System (INIS)

    Chernyavskaya, N.E.; Ochkin, A.V.; Chizhevskaya, S.V.; Stefanovskij, S.V.

    1998-01-01

    SYNROC is titanate ceramics consisting mainly of zirconolite, perovskite, and hollandite, developed to immobilize high level waste. Perovskite is able to incorporate strontium, yttrium, and trivalent lanthanides and actinides. The main goal of the present work is leaching study of various radionuclides from perovskite. Samples of perovskite-rich ceramics were produced by cold pressing of oxide mixture followed by firing in resistive furnace at 1350 degC for 3 hours. For leaching tests, ceramic pellets were crushed and surface areas were measured using argon thermal desorption technique. Leach rate was measured by boiling in a Soxhlet apparatus for 5 hours. Leach rates in 0.1 M HNO 3 and NaCl solutions were measured by boiling with stirrer and reverse cooler. Leach rate was controlled with radioactive indicator technique. Density of the perovskite-rich ceramic samples prepared was about 75% of theoretical. From XRD examination, the target phase (perovskite) yield was found to be about 95 vol.%. Minor rutile (≤ 5 vol.%) was also present. Leach rate of 90 Sr from Sr-doped perovskites with specified composition Ca 1-x Sr x TiO 3 did not depend on x until certain x value. Leach rate of 90 Sr from control zirconolite sample was by one order of magnitude higher than from perovskite. Leach rates of 147 Pm, 238 Pu, and 241 Am from perovskite ceramics with nominal perovskite composition had the same order of magnitude (about 10 -4 g/(m 2 day)). Substitution of 5 at.% Ce for Ca and 5 at.% Al for Ti lowered leach rate of 238 Pu by a factor of 6. Leach rates of 90 Sr in 0.1 M HNO 3 and NaCl solutions were three and one orders of magnitude higher than in distilled water

  4. Solution based synthesis of perovskite-type oxide films and powders

    International Nuclear Information System (INIS)

    McHale, J.M. Jr.

    1995-01-01

    Conventional solid state reactions are diffusion limited processes that require high temperatures and long reaction times to reach completion. In this work, several solution based methods were utilized to circumvent this diffusion limited reaction and achieve product formation at lower temperatures. The solution methods studied all have the common goal of trapping the homogeneity inherent in a solution and transferring this homogeneity to the solid state, thereby creating a solid atomic mixture of reactants. These atomic mixtures can yield solid state products through diffusionless mechanisms. The effectiveness of atomic mixtures in solid state synthesis was tested on three classes of materials, varying in complexity. A procedure was invented for obtaining the highly water soluble salt, titanyl nitrate, TiO(NO 3 ) 2 , in crystalline form, which allowed the production of titanate materials by freeze drying. The freeze drying procedures yielded phase pure, nanocrystalline BaTiO 3 and the complete SYNROC-B phase assemblage after ten minute heat treatments at 600 C and 1,100 C, respectively. Two novel methods were developed for the solution based synthesis of Ba 2 YCu 3 O 7-x and Bi 2 Sr 2 Ca 2 Cu 3 O 10 . Thin and thick films of Ba 2 YCu 3 O 7-x and Bi 2 Sr 2 Ca 2 Cu 3 O 10 were synthesized by an atmospheric pressure, chemical vapor deposition technique. Liquid ammonia solutions of metal nitrates were atomized with a stream of N 2 O and ignited with a hydrogen/oxygen torch. The resulting flame was used to coat a substrate with superconducting material. Bulk powders of Ba 2 YCu 3 O 7-x and Bi 2 Sr 2 Ca 2 Cu 3 O 10 were synthesized through a novel acetate glass method. The materials prepared were characterized by XRD, TEM, SEM, TGA, DTA, magnetic susceptibility and electrical resistivity measurements

  5. Emerging materials by advanced processing

    International Nuclear Information System (INIS)

    Kaysser, W.A.; Weber-Bock, J.

    1989-01-01

    This volume contains 36 contributions with following subjects (selection): Densification of highly reactive aluminium titanate powders; influence of precursor history on carbon fiber characteristics; influence of water removal rate during calcination on the crystallization of ZrO2 from amorphous hydrous precipitates; tape casting of AlN; influence of processing on the properties of beta-SiC powders; corrosion of SiSiC by gases and basic slag at high temperature; influence of sintering and thermomechanical treatment on microstructure and properties of W-Ni-Fe alloys; mechanical alloying for development of sintered steels with high hard phase content (NbC); early stages of mechanical alloying in Ni-Ti and Ni-Al powder mixtures; growth and microstructural development of melt-oxidation derived Al2O3/Al-base composites; fabrication of RSBN composites; synthesis of high density coridierite bodies; comparative studies on post-HIP and sinter-HIP treatments on transformation thoughened ceramics; sinter HIP of SiC; precipitation mixing of Si3N4 with bimetallic oxides; temperature dependence of the interfacial energies in Al2O3-liquid metal systems; synthesis and microstructural examination of Synroc B; solid state investigation of ceramic-metal bonding; thermophysical properties of MgAl2O4; preparation, sintering and thermal expansion of MgAl2O4; microstructural studies on alumina-zirconia and metallized alumina ceramics; electrodeposition of metals (e.g. Ti, Mo, In) and metal oxides from molten salts; electrochemical deposition of Ti from nonaqueous media (DMSO, DMF); lithium as anode material in power sources (passivation); reduction of chromium(VI) when solar selective black chromium is deposited; thermodynamic optimization of phase diagrams (computer calculations); optimization of Na-Tl phase diagram; phase relations in the Y-Si-Al-O-N system: Controlled manufacturing of alpha/beta-SIALON composites. (MM)

  6. Radionuclide separations in the nuclear fuel cycle development and application of micro and meso porous inorganic ion-exchangers

    International Nuclear Information System (INIS)

    Griffith, C.S.; Luca, V.

    2006-01-01

    Full text: Full text: From the mining of uranium-containing ores to the reprocessing of spent nuclear fuel, separations technologies play a crucial role in determining the efficiency and viability of the nuclear fuel cycle. With respect to proposed Advanced Nuclear Fuel Cycles (ANFC), the integral role of separations is no different with solvent extraction and pyroelectrometalurgical processing dominating efforts to develop a sustainable and publicly acceptable roadmap for nuclear power in the next 100 years. An often forgotten or overlooked separation technology is ion-exchange, more specifically, inorganic ion-exchangers. This is despite the fact that these materials offer the potential advantages of process simplicity; exceptional selectivity against high background concentrations of competing ions; and the possibility of a simple immobilization route for the separated radionculides. ANSTO's principal interest in inorganic ion-exchange materials in recent years has been the development of an inorganic ion-exchanger for the pretreatment of acidic legacy 9 Mo production waste to simultaneously remove radiogenic cesium and strontium. Radiogenic cesium and strontium comprise the majority of activity in such waste and may offer increased ease in the downstream processing to immobilise this waste in a Synroc wasteform. With the reliance on separations technologies in all current ANFC concepts, and the recent admission of ANSTO to the European Commissions EUROPART project, the development of new inorganic ion-exchangers has also expanded within our group. This presentation will provide a background of the fundamentals of inorganic and composite inorganic-organic ion-exchange materials followed by specific discussion of some selected inorganic and composite ion-exchange materials being developed and studied at ANSTO. The detailed structural and ion-exchange chemistry of these materials will be discussed and note made of how such materials could benefit any of the

  7. 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

  8. 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)

  9. Long-term criticality safety concerns associated with surplus fissile material disposition

    International Nuclear Information System (INIS)

    Choi, J.S.

    1995-01-01

    A substantial inventory of surplus fissile material would result from ongoing and planned dismantlement of US and Russian nuclear weapons. This surplus fissile material could be dispositioned by irradiation in nuclear reactors, and the resulting spent MOx fuel would be similar in radiation characteristics to regular LWR spent UO2 fuel. The surplus fissile material could also be immobilized into high-level waste forms, such as borosilicate glass, synroc, or metal-alloy matrix. The MOx spent fuel, or the immobilized waste forms, could then be directly disposed of in a geologic repository. Long-term criticality safety concerns arise because the fissile contents (i.e., Pu-239 and its decay daughter U-235) in these waste forms are higher than in LWR spent UO2 fuel. MOx spent fuel could contain 3 to 4 wt% of reactor-grade plutonium, compared to only 0.9 wt% of plutonium in LWR spent UO2 fuel. At some future time (tens of thousand of years), when the waste forms had deteriorated due to intruding groundwater, the water could mix with the long-lived fissile materials to form into a critical system. If the critical system is self-sustaining, somewhat like the natural-occurring reactor in OKLO, fission products produced could readily be available for dissolution and release out to the accessible environment, adversely affecting public health and safety. This paper will address ongoing activities to evaluate long-term criticality safety concerns associated with disposition of fissile material in a geologic setting. Issues to be addressed include the identification of a worst-case water-intrusion scenario and waste-form geometries which present the most concern for long-term criticality safety; and suggests of technical solutions for such concerns

  10. Use of linear free energy relationship to predict Gibbs free energies of formation of zirconolite phases (MZrTi2O7 and MHfTi2O7)

    International Nuclear Information System (INIS)

    Xu, H.

    1999-01-01

    In this letter, the Sverjensky-Molling equation derived from a linear free energy relationship is used to calculate the Gibbs free energies of formation of zirconolite crystalline phases (MZrTi 2 O 7 and MHfTi 2 O 7 ) from the known thermodynamic properties of the corresponding aqueous divalent cations (M 2+ ). Sverjensky-Molling equation is expressed as ΔG 0 f,M v X =a M v X ΔG 0 n,M 2+ +b M v X +β M v X r M 2+ , where the coefficients a M v X , b M v X , and β M v X characterize a particular structural family of M v X, r M 2+ is the ionic radius of M 2+ cation, ΔG f,M v X 0 is the standard Gibbs free energy of formation of M v X, and ΔG 0 n,M 2+ is the standard non-solvation energy of cation M 2+ . This relationship can be used to predict the Gibbs free energies of formation of various fictive phases (such as BaZrTi 2 O 7 , SrZrTi 2 O 7 , PbZrTi 2 O 7 , etc.) that may form solid solution with CaZrTi 2 O 7 in actual Synroc-based nuclear waste forms. Based on obtained linear free energy relationships, it is predicted that large cations (e.g., Ba and Ra) prefer to be in perovskite structure, and small cations (e.g., Ca, Zn, and Cd) prefer to be in zirconolite structure. (orig.)

  11. Volatilization of heavy metals and radionuclides from soil heated in an induction ''cold'' crucible melter

    International Nuclear Information System (INIS)

    Aloy, A.S.; Belov, V.Z.; Trofimenko, A.S.; Dmitriev, S.A.; Stefanovsky, S.V.; Gombert, D.; Knecht, D.A.

    1997-01-01

    The behavior of heavy metals and radionuclides during high-temperature treatment is very important for the design and operational capabilities of the off-gas treatment system, as well as for a better understanding of the nature and forms of the secondary waste. In Russia, a process for high-temperature melting in an induction heated cold crucible system is being studied for vitrification of Low Level Waste (LLW) flyash and SYNROC production with simulated high level waste (HLW). This work was done as part of a Department of Energy (DOE) funded research project for thermal treatment of mixed low level waste (LLW). Soil spiked with heavy metals (Cd, Pb) and radionuclides (Cs-137, U-239, Pu-239) was used as a waste surrogate. The soil was melted in an experimental lab-scale system that consisted of a high-frequency generator (1.76 MHz, 60 kW), a cold crucible melter (300 mm high and 90 mm in diameter), a shield box, and an off-gas system. The process temperature was 1,350--1,400 C. Graphite and silicon carbide were used as sacrificial conductive materials to start heating and initial melting of the soil batch. The off-gas system was designed in such a manner that after each experiment, it can be disconnected to collect and analyze all deposits to determine the mass balance. The off-gases were also sampled during an experiment to analyze for hydrogen, NO x , carbon dioxide, carbon monoxide and chlorine formation. This paper describes distribution and mass balance of metals and radionuclides in various parts of the off-gas system. The leach rate of the solidified blocks identified by the PCT method is also reported

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

    International Nuclear Information System (INIS)

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

    1980-09-01

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

  13. Solution based preparation of Perovskite-type oxide films and powders

    Energy Technology Data Exchange (ETDEWEB)

    McHale, Jr., James M. [Temple Univ., Philadelphia, PA (United States). Dept. of Chemistry

    1995-04-01

    Conventional solid state reactions are diffusion limited processes that require high temperatures and long reaction times to reach completion. In this work, several solution based methods were utilized to circumvent this diffusion limited reaction and achieve product formation at lower temperatures. The solution methods studied all have the common goal of trapping the homogeneity inherent in a solution and transferring this homogeneity to the solid state, thereby creating a solid atomic mixture of reactants. These atomic mixtures can yield solid state products through "diffusionless" mechanisms. The effectiveness of atomic mixtures in solid state synthesis was tested on three classes of materials, varying in complexity. A procedure was invented for obtaining the highly water soluble salt, titanyl nitrate, TiO(NO3)2, in crystalline form, which allowed the production of titanate materials by freeze drying. The freeze drying procedures yielded phase pure, nanocrystalline BaTiO3 and the complete SYNROC-B phase assemblage after ten minute heat treatments at 600{degrees}C and 1100{degrees}C, respectively. Two novel methods were developed for the solution based synthesis of Ba2YCu3O7-x and Bi2Sr2Ca2Cu3O10. Thin and thick films of Ba2YCu3O7-x and Bi2Sr2Ca2u3O10 were synthesized by an atmospheric pressure, chemical vapor deposition technique. Liquid ammonia solutions of metal nitrates were atomized with a stream of N2O and ignited with a hydrogen/oxygen torch. The resulting flame was used to coat a substrate with superconducting material. Bulk powders of Ba2YCu3O7-x and Bi2Sr2Ca2Cu3O10 were synthesized through a novel acetate glass method. The materials prepared were

  14. Ion irradiation of rare-earth- and yttrium-titanate-pyrochlores

    International Nuclear Information System (INIS)

    Wang, S.X.; Wang, L.M.; Ewing, R.C.; Govindan Kutty, K.V.

    2000-01-01

    Pyrochlore, A 1-2 B 2 O 6 (O,OH,F) 0-1 , is an actinide-bearing phase in Synroc, a polyphase ceramic proposed for the immobilization of high level nuclear waste. Structural damage due to alpha-decay events can significantly affect the chemical and physical stability of the nuclear waste form. Pyrochlore can effectively incorporate a variety of actinides into its structure. Four titanate pyrochlores were synthesized with compositions of Gd 2 Ti 2 O 7 , Sm 2 Ti 2 O 7 , Eu 2 Ti 2 O 7 and Y 2 Ti 2 O 2 . These samples were irradiated with 1 MeV Kr + in order to simulate alpha-decay damage and were observed by in situ electron microscopy. Irradiations were conducted from 25 K to 1023 K. At room temperature, Gd-, Sm- and Eu-pyrochlores amorphized at a dose of ∼2x10 14 ions/cm 2 (∼0.5 dpa) and Y-pyrochlore amorphized at 4x10 14 ions/cm 2 (∼0.8 dpa). The amorphization dose became higher at elevated temperatures with different rates of increase for each composition. The critical temperatures for amorphization are ∼1100 K for Gd-, Sm-, Eu-pyrochlore and ∼780 K for Y-pyrochlore. The rare-earth-pyrochlores are more susceptible to amorphization and have higher critical temperatures than Y-pyrochlore. The difference in amorphization dose and critical temperature is attributed to the different cascade sizes caused by the different cation masses of the target. Based on a model of cascade quenching, the larger cascade is related to a lower amorphization dose and higher critical temperature. The irradiated materials were studied by electron diffraction and high-resolution electron microscopy. All the pyrochlores transformed to a fluorite substructure prior to the completion of amorphization of the observed regions. This transformation was caused by the disordering between cations and between oxygen and oxygen vacancies. The concurrence of cation disordering with amorphization suggests the partial recrystallization of the displacement cascades. Isolated cascade damage

  15. A novel hydrothermal method to convert incineration ash into pollucite for the immobilization of a simulant radioactive cesium

    Energy Technology Data Exchange (ETDEWEB)

    Jing, Zhenzi, E-mail: zzjing@tongji.edu.cn [Key Laboratory of Advanced Civil Engineering Materials, Ministry of Education, Tongji University, 4800 Cao’an Road, Shanghai 201804 (China); Hao, Wenbo; He, Xiaojun; Fan, Junjie; Zhang, Yi; Miao, Jiajun [Key Laboratory of Advanced Civil Engineering Materials, Ministry of Education, Tongji University, 4800 Cao’an Road, Shanghai 201804 (China); Jin, Fangming [School of Environmental Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China)

    2016-04-05

    Highlights: • Incineration ash could be converted hydrothermally to pollucite to immobilize Cs. • Pollucite could be synthesized readily with a wide range of Cs/Si ratios. • With Ca(OH){sub 2} added, a tough pollucite could be used to solidify Cs-polluted RHA. • Leaching results showed that the amount of Cs leached from specimen was very low. - Abstract: The Fukushima nuclear accident in Japan on March 11, 2011 produced huge amounts of Cs-polluted incineration ashes; conventional solidification methods seem unsuitable for the treatment of large amounts of Cs-polluted ashes. A novel hydrothermal method was developed to directly convert Cs-polluted incineration ash (rice husk ash) into pollucite to immobilize Cs in its crystal structure in situ. Results revealed that pollucite could be synthesized readily over a wide range of added Cs (Cs/Si = 0.2–0.6); the addition of more Cs (Cs/Si ≥ 0.5) caused the formation of a small amount of cesium aluminosilicate (CsAlSiO{sub 4}), which exhibits poor immobilization behavior for Cs. Pollucite could be formed even for a short curing time (1 h) or at a low curing temperature (150 °C). However, a high curing temperature or a long curing time favored the formation of a pure pollucite. With the added calcium hydroxide, a tough specimen with a flexural strength of approximately 22 MPa could be obtained, which suggested that this technology may be applied directly to the solidification of Cs-polluted incineration ashes. Hydrogarnet and tobermorite formations enhanced the strength of the solidified specimens, and meanwhile the formed pollucite was present in a matrix steadily. Leaching test demonstrated that the amount of Cs that leached from the synthesized specimens was very low (0.49 × 10{sup −5}–2.31 × 10{sup −5}) and even lower than that from the reference hollandite-rich synroc (2.0 × 10{sup −2}), although a higher content of Cs was found in the synthesized pollucite specimens (6.0–31.7%) than in the

  16. Pyrochlore as nuclear waste form. Actinide uptake and chemical stability

    Energy Technology Data Exchange (ETDEWEB)

    Finkeldei, Sarah Charlotte

    2015-07-01

    Radioactive waste is generated by many different technical and scientific applications. For the past decades, different waste disposal strategies have been considered. Several questions on the waste disposal strategy remain unanswered, particularly regarding the long-term radiotoxicity of minor actinides (Am, Cm, Np), plutonium and uranium. These radionuclides mainly arise from high level nuclear waste (HLW), specific waste streams or dismantled nuclear weapons. Although many countries have opted for the direct disposal of spent fuel, from a scientific and technical point of view it is imperative to pursue alternative waste management strategies. Apart from the vitrification, especially for trivalent actinides and Pu, crystalline ceramic waste forms are considered. In contrast to glasses, crystalline waste forms, which are chemically and physically highly stable, allow the retention of radionuclides on well-defined lattice positions within the crystal structure. Besides polyphase ceramics such as SYNROC, single phase ceramics are considered as tailor made host phases to embed a specific radionuclide or a specific group. Among oxidic single phase ceramics pyrochlores are known to have a high potential for this application. This work examines ZrO{sub 2} based pyrochlores as potential nuclear waste forms, which are known to show a high aqueous stability and a high tolerance towards radiation damage. This work contributes to (1) understand the phase stability field of pyrochlore and consequences of non-stoichiometry which leads to pyrochlores with mixed cationic sites. Mixed cationic occupancies are likely to occur in actinide-bearing pyrochlores. (2) The structural uptake of radionuclides themselves was studied. (3) The chemical stability and the effect of phase transition from pyrochlore to defect fluorite were probed. This phase transition is important, as it is the result of radiation damage in ZrO{sub 2} based pyrochlores. ZrO{sub 2} - Nd{sub 2}O{sub 3} pellets

  17. Pyrochlore as nuclear waste form. Actinide uptake and chemical stability

    International Nuclear Information System (INIS)

    Finkeldei, Sarah Charlotte

    2015-01-01

    Radioactive waste is generated by many different technical and scientific applications. For the past decades, different waste disposal strategies have been considered. Several questions on the waste disposal strategy remain unanswered, particularly regarding the long-term radiotoxicity of minor actinides (Am, Cm, Np), plutonium and uranium. These radionuclides mainly arise from high level nuclear waste (HLW), specific waste streams or dismantled nuclear weapons. Although many countries have opted for the direct disposal of spent fuel, from a scientific and technical point of view it is imperative to pursue alternative waste management strategies. Apart from the vitrification, especially for trivalent actinides and Pu, crystalline ceramic waste forms are considered. In contrast to glasses, crystalline waste forms, which are chemically and physically highly stable, allow the retention of radionuclides on well-defined lattice positions within the crystal structure. Besides polyphase ceramics such as SYNROC, single phase ceramics are considered as tailor made host phases to embed a specific radionuclide or a specific group. Among oxidic single phase ceramics pyrochlores are known to have a high potential for this application. This work examines ZrO 2 based pyrochlores as potential nuclear waste forms, which are known to show a high aqueous stability and a high tolerance towards radiation damage. This work contributes to (1) understand the phase stability field of pyrochlore and consequences of non-stoichiometry which leads to pyrochlores with mixed cationic sites. Mixed cationic occupancies are likely to occur in actinide-bearing pyrochlores. (2) The structural uptake of radionuclides themselves was studied. (3) The chemical stability and the effect of phase transition from pyrochlore to defect fluorite were probed. This phase transition is important, as it is the result of radiation damage in ZrO 2 based pyrochlores. ZrO 2 - Nd 2 O 3 pellets with pyrochlore and defect