WorldWideScience

Sample records for synroc

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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

  17. SYNROC powder preparation: preliminary fluid bed tests in a cold-flow unit

    International Nuclear Information System (INIS)

    Peters, P.E.; Ackerman, F.J.; Grens, J.Z.

    1982-01-01

    SYNROC is a titanate-based material which is being developed as a medium for immobilizing high-level nuclear waste. The use of a fluid bed unit for production of SYNROC powder has a precedent in the use of a fluid-bed calciner for high-level waste processing at the Idaho Chemical Processing Plant (Idaho Falls, ID). In order to facilitate the design of a fluid bed and demonstrate its use for SYNROC production, two small units have been constructed, one for low temperature use and a second for actual SYNROC production at temperatures up to 800 0 C. The low-temperature unit is constructed with glass walls to allow observation of the fluidization process, including the onset of fluidization, bed and gas bubble behavior and mixing phenomena. Disturbances caused by side streams entering the bed have been examined. Side streams may represent fuel and oxidizer admission or slurry feed with subsequent flash vaporization of the carrier liquid. This report is a summary of the initial tests made with the low-temperature, glass-walled fluid bed. The tests described include: (1) the measurement of basic fluidization parameters; (2) measurement of the effect of distribution-plate design on bed fluidization; (3) observation of jet penetration, bubble formation and coalescence, and surface behavior; and (4) studies of mixing within the bed

  18. Influence of melting atmosphere on Synroc-C microstructure and phase composition

    Energy Technology Data Exchange (ETDEWEB)

    Day, R.A.; La Robina, M.; Moricca, S.; Eddowes, T.; Blagojevic, N.; Carter, L. [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia)

    1999-07-01

    In this study, comparable batches of Synroc were melted in air and in a reducing atmosphere consisting of 3% hydrogen in nitrogen. The material used for both the air and the H{sub 2}/N{sub 2} atmosphere consisted of about 500 gram batches of Synroc-C with 20 wt% added PW-4b simulated Purex-type radwaste and 2 wt% added Ti metal. This material had previously been hot pressed under reducing conditions. Thefollowing results are obtained. Melting in either atmosphere resulted in the crystallization of all of the major Synroc titanate phases. Melts produced in both atmospheres resulted in less zirconolite than in the hot pressed equivalent, while the air melt had less perovskite and the H{sub 2}/N{sub 2} melt had sightly more perovskite than the hot pressed equivalent. Zirconolite was larger and better formed in the air melt than in the H{sub 2}/N{sub 2} melt where it was more dendritic and often associated with perovskite. Hollandite was generally larger and better formed in the H{sub 2}/N{sub 2} melt. The main consequences of melting in air were that basically all of the caesium, as well as some of the barium, strontium and calcium were diverted into less chemically stable non-Synroc phases such as molybdates and phosphates instead of into the intended hollandite and perovskite. In the hot pressed Synroc-C the intended hosts for molybdenum and phosphorus are metallic molybdenum alloys and phosphides while all of the caesium and barium enters hollandite. When the Synroc was melted in an H{sub 2}/N{sub 2} atmosphere, the outcome was broadly the same as for hot pressed Synroc, namely molybdenum dominated metallic phases which also incorporated the phosphorus, while caesium and the rest of barium entered hollandite. Also the rest of the strontium and calcium entered perovskite. The conclusion, based on the above comparison, is that it is possible to use the melting atmosphere to optimize the oxidation states of species in cold crucible melted ceramics. Control of the

  19. 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

  20. 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

  1. 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

  2. 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

  3. Investigations of the volatilization of molybdenum and ruthenium during subsolidus sintering of modified SYNROC-B crystalline waste forms

    International Nuclear Information System (INIS)

    Solomah, A.G.; Odoj, R.

    1984-01-01

    Volatilizations of molybdenum and ruthenium during the fixation of simulated high-level radioactive waste in modified SYNROC-B crystalline ceramic waste forms have been studied using a radiotracer technique. The simulated waste loading was 20 wt%. The volatilization figures of merit (VFMs) for 99 Mo and 103 Ru have shown a behavior that depends on the type of sintering atmosphere, i.e., oxidizing versus reducing. The experimentally obtained VFM /SUB Mo/ and VFM /SUB Ru/ in an oxidizing atmosphere are 7.8 and 3.7% of the initial radioactivity of each nuclide per gram of sintered SYNROC-B product after sintering at 1510 K in air, while under reducing conditions (50% H 2 -50% Ar), VFM /SUB Mo/ and VFM /SUB Ru/ have been reduced to 2.8 and 1.8% g -1 , respectively. Solidification of high-level radioactive waste in the proposed waste form or in glass matrices under reducing atmosphere is recommended to minimize the amounts of volatilization and, subsequently, to reduce the safety requirements of the off-gas treatment system of the vitrification and/or solidification plant

  4. 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.

  5. 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

  6. Altération en présence d'argile humide à 70°C de céramiques à base de Ti et Zr, de type Synroc

    Science.gov (United States)

    Leturcq, Gilles; Berger, Gilles; Advocat, Thierry; Lacombe, Jacques; Vance, Eric

    1998-12-01

    The chemical durability of three zirconotitanate ceramics was quantified on contact with humid clay at 70°C. These materials are innovative candidates for the containment of fission products or long-lived actinides such as Np, Am and Pu. Synroc C containing molybdates is the only sample showing significant mass losses after nine months of interaction with humid clay. This may be due to a migration of Mo to the reaction interface. Mo was recovered in clay as calcium molybdate.

  7. 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

  8. 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

  9. 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)

  10. 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

  11. New immobilisation methods for radioactive waste. Metal composite and other systems

    International Nuclear Information System (INIS)

    Ozhovan, M.

    2004-01-01

    New immobilisation hosts and technologies are presented. Some new approaches as crystalline hosts, polyphase crystalline forms (SYNROC), polyphase forms (composites), metal matrix immobilisation are discussed. The potential use and chemical properties and radiation durability of minerals Monazite, Zircon and Zirconolite, Hollandite, Apatites, Britolite and NZP are presented. The most famous polyphase ceramic for nuclear waste immobilisation is SYNROC. The properties of SYNROC and a comparison of SYNROC matrix parameters with nuclear waste glasses is made. Glass composites may be used to immobilise long-lived radionuclides (e.g. An) by incorporating them into the more durable crystalline phases, whereas the short-lived radionuclides may be accommodated in the less durable vitreous phase. An example of such glass composite is so-called SYNROC-glass, which is a glass-composite material with SYNROC crystalline phases in a vitreous matrix. The new technological approaches discussed in the paper are: melting, sintering, thermochemical method. The features and advantages of metal matrix immobilization are also discussed

  12. New ceramic prevents atomic waste leaks, Australian scientists say

    International Nuclear Information System (INIS)

    1991-01-01

    Scientists at the Australian National University say they have invented a material that will safely seal radioactive atomic waste and prevent leaks from occurring for millions for years. Synroc, or synthetic rock, is an advanced ceramic which, when fused with radioactive waste, locks the waste in a crystalline structure, according to a Reuters report. A five kilogram chunk of chemically-inactive synroc is all that is left after 100 litres of radioactive waste is processed. The ceramic was developed in conjunction with the Australian Nuclear Science and Technology Organization and a consortium of four mining companies

  13. Bulletin of Materials Science | Indian Academy of Sciences

    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.

  14. 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

  15. 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 ...

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

    Indian Academy of Sciences (India)

    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 Ca1-TiO3 (where = 0.05) has been synthesized through ceramic route by taking calculated quantities of oxides of Ca, Ti and La as ...

  17. 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

  18. Phase transformations in the Ca1xSrxTiO3 perovskite system

    International Nuclear Information System (INIS)

    Daniels, J.E.; Finlayson, T.R.; Elcombe, M.M.; Vance, E.R.

    2004-01-01

    Full text: The mixed perovskite Ca 1-x Sr x TiO 3 is one of the major constituent phases of SYNROC, a synthetic rock form proposed for the long term immobilisation of high-level nuclear waste. A complete understanding of the crystal structures formed, and transitions within, each of the constituent phases is of vital importance to gaining a complete knowledge of the overall structural stability of SYNROC. This study investigated the transition temperatures and the space group symmetry of several samples between the composition values of x = 0.6 to 0.85. Previous studies of this compound have produced many conflicting results. It is believed that the reason for this is the sample preparation technique coupled with the form of the sample during experimentation, i.e., powder vs. polycrystalline solid. This study has attempted to produce samples using techniques closely related to those used in bulk SYNROC manufacture and to analyse these samples in their polycrystalline form as they would exist in SYNROC. Transition temperatures were determined by observing the dynamic Young's modulus and internal friction of the samples between 4K and 420K using the Piezoelectric Ultrasonic Composite Oscillator Technique. Classifications of the crystal structures formed were carried out using the High Resolution Powder Diffractometer at the Lucas Heights Research Reactor within the temperature range of 8K to 300K. Space group symmetries for each phase were then determined by Rietveld refinement

  19. 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.

  20. Nuclear waste immobilization. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Ringwood, A.E.; Sinclair, W.; McLaughlin, G.M.

    1979-11-20

    United States defense nuclear wastes are presently in tank storage, largely as sludges comprising Fe, Mn, Ni, U and Na oxides and hydroxides, together with 0.5 to 5 percent of fission products and actinides (exclusive of uranium). The relative proportions of Al, Fe, Mn, Ni, U and Na in the sludges from different tanks vary considerably, except that (Fe + Al + Mn) are by far the major components and Fe is more abundant than Mn. Typical compositions of some calcined sludges from Savannah River are given. This paper briefly describes how the SYNROC process, utilizing straightforward technology, can be readily adapted to the problem of defense waste immobilization, yielding a dense, inert, ceramic waste-form, SYNROC-D. Two classes of processes are discussed - one designed to immobilize sludges containing normal amounts of sodium and the other designed for otherwise similar sludges which are, however, strongly depleted in sodium as a result of more efficient washing procedures.

  1. Three-dimensional thermal analysis of in-floor type nuclear waste repository for a ceramic waste form

    International Nuclear Information System (INIS)

    Sizgek, G. Devlet

    2005-01-01

    A thermal model is constructed and analyses are performed for an 'in-floor' type nuclear waste repository in granitic rock for a high level nuclear waste (HLW)-bearing ceramic waste form (synroc). Transient calculations for a three-dimensional (3-D) model have been carried out for both 20 and 10 wt.% HLW-bearing synroc, for surface cooling periods between reactor discharge and geological disposal varying from 5 to 40 years. This study investigates the temperature distribution in one of the boreholes of a hypothetical tunnel for a basic geometrical setting as well as the effect of varying the distance between adjacent boreholes and the distance between adjacent tunnels. The temperatures in the repository were found to be sensitive to the interim surface cooling period as well as the amount of waste loaded. The results showed that decreasing the spacing between the canisters has a more pronounced effect on the temperature field than decreasing the spacing between the tunnels

  2. 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

  3. Perovskite synthesis via complex sol-gel process to immobilize radioactive waste elements

    International Nuclear Information System (INIS)

    Smolinski, T.; Deptula, A.; Olczak, T.; Lada, W.; Brykala, M.; Wojtowicz, P.; Wawszczak, D.; Rogowski, M.; Zaza, F.

    2014-01-01

    Synroc (Synthetic Rocks) materials have been regarded as the second generation of high level waste forms in the world. It allows incorporating into their crystal structures almost all of the elements present in high-level radioactive waste. One of the components of Synroc-C is perovskite (CaTiO 3 ) which immobilize mainly fission products, but also allow immobilizing in his structure long-lived actinides such as plutonium (Pu). Perovskite phase has been fabricated by a sol-gel route. In the present work complex sol-gel process (CSGP Polish Patent PL 172618, 1997) and method of synthesis Me-titanates (Polish Patent PL 198039, 2001) were adapted to prepare of perovskite. Additions of 10 % molar Sr, Co, Cs and Nd into Ti-Ca-nitrate sols were carried out by CSGP. Gels obtained by evaporation of sols under reduced pressure were thermal treated according thermogravimetric (TG, DTA) analysis. Transformation of ascorbat-nitrate gels into doped orthorhombic perovskite phases was definitely lower (about 650 deg C) than those pure nitrate gels (approx. 700 deg C). All structures were confirmed by X-ray diffraction analyses. Surrogates were homogeneously distributed into crystalline structure of perovskite. It means that elaborated process can be applied for synthesis Synroc materials and it might be competitive to vitrification process. (author)

  4. 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

  5. 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

  6. 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

  7. Young Investigator Program: Quasi-Liquid Grain Boundary Films in Refractory Metals

    Science.gov (United States)

    2010-01-15

    Ga [72-75] Corrosion of synroc [76] GB embrittlement of W-Ni, Cu-Bi, and other systems [42] Solid-State Activated Sintering ZnO-Bi2O3; CeO2 ...nickel in molybdenum, nickel chloride (NiCl2·6H2O, Alfa Aesar) was mixed with molybdenum in a solution/suspension. Then, the slurry was dried in an oven...used to measure the grain size. The specimens were polished to mirror finish on emery papers and alumina slurries (or diamond paste). Thereafter

  8. 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

  9. 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

  10. 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

  11. 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

  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. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

    Borosilicate Glass Marbles and/or monoliths were rated among the top three waste forms for immobilization of all types of Hanford high-level waste. Supergrout Concrete and Bitumen, low temperature processes, are judged to be particularly suitable for immobilization and bulk disposal of high sodium blended wastes and/or residual liquid. This preliminary assessment indicates that certain ceramic waste forms (e.g., Tailored Ceramics, Supercalcine Ceramic, and SYNROC Ceramic) are equal to or superior to Borosilicate Glass waste forms for immobilization of Hanford sludges and radionuclides removed from salt cake and residual liquid. These ceramic waste forms can be made by the Sol Gel process. Some multibarrier waste forms (e.g., Coated Ceramics, Ceramic Pellets in Metal Matrix, and Glass in Metal Matrix) are judged to be superior waste forms for immobilization of Hanford sludges and/or radionuclide concentrate

  1. 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)

  2. ANSTO at work

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    The Australian Nuclear Science and Technology Organization (ANSTO) has world recognized capabilities in nuclear science and technology. This paper describes some of the applications of the radioisotopes produced at ANSTO, within the fields of medicine, agriculture, industry and research. Radioisotope Tracer Technology, involving Technetium 99, Copper 64, Indium 111 or Gallium 67 are becoming widely used for early diagnosis of conditions such as cancer, coronary disease, strokes and severe trauma. As a result of collaborative research between ANSTO and two Sydney hospitals, a fairly accurate measure of total body protein has been developed, based on the capture of neutrons by nitrogen nuclei. The National Medical Cyclotron, a new facility expected to be operational by 1991, will give Australian medical technologists access to a much wider variety of radioisotope tracers. Other activities involved in product or process developments briefly described include estuary siltation, soil erosion and salinity problems, the controlled doping of semiconductor silicon and Synroc process. ills

  3. 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

  4. 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

  5. Progress report on safety research of high-level waste management for the period April 1985 to March 1986

    International Nuclear Information System (INIS)

    Nakamura, Haruto; Tashiro, Shingo

    1986-09-01

    Researches on high level waste management in fiscal year of 1985 is reviewed. Topics are as follows; 1) Glass waste form was examined with emphasis on the leaching mechanisms under various conditions to predict the long-term leach rates. Leaching rate was examined in synthesized groundwater and a leaching model was developed. Cooperation between Japan and Australia on development of SYNROC has started. 2) Heating experiments with a real size simulated canister and migration tests using non-sorbing tracer has been carried out in a near surface granite rock mass. 2D-SEEP, the coupled computer code of heat and groundwater flow, has been developed. 3) Japanese group participated the ESOPE project of OECD/NEA SWG. Small particles of an organic phase was found in sediment as a material to have significant influence on technetium fixation. 4) Alpha radiation stability of vitrified forms under beta and gamma irradiation have newly started in WASTEF. (author)

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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

  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. Support for DOE program in mineral waste-form development

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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)

  2. 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

  3. 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

  4. Structural effect of actinide substitutions on the Ca and Zr sites of zirconolite

    International Nuclear Information System (INIS)

    Begg, B.; Vance, E.; Conradson, S.

    2000-01-01

    Full text: Recently a SYNROC derivative was chosen as the wasteform for the immobilisation of surplus plutonium in the US. The major phase for the immobilisation of the plutonium is a titanate-based pyrochlore. Pyrochlore is stabilised when rare-earth or actinide substitutions in excess of 0.5 formula units are attempted on the Zr site of zirconolite. Substitutions of between 0.15 and 0.5 formula units on the Zr site are known to stabilise the zirconolite-4M polytype. This paper examines both the short and long-range structural impact of substituting Pu and U on the Zr site of zirconolite, leading to the stabilisation of pyrochlore. The impact of tri- and tetravalent Pu substitutions on the Ca site of zirconolite will also be discussed. Pu LIII- and Zr K-edge extended X-ray absorption fine structure (EXAFS) were used to characterise the short-range structures around both the substituted Pu and the Zr host lattice site, whilst X-ray diffraction provided long-range structural information. Copyright (2000) American Chemical Society

  5. 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

  6. 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

  7. Present status of the Waste Safety Testing Facility

    International Nuclear Information System (INIS)

    Kikuchi, A.; Yamada, N.

    1993-01-01

    The Waste Safety Testing Facility (WASTEF) was established in 1981, in which the objective was to evaluate the confinement performance of glass and Synroc waste forms including high level waste (HLW). To this target, the following examinations have been typically carried out ; the fabrication and characterization of waste forms, the volatility test as a storage behavior and the leachability test as a disposal behavior. The facility is composed of three beta/gamma concrete cells, two alpha/gamma concrete cells, one lead cell and five glove boxes. The lead cell and glove boxes are of alpha/gamma type and attached to the alpha/gamma concrete cells No. 4 and No. 5, respectively. Several kinds of testing apparatus, measuring instruments and analytical equipments are located in the cells, the glove boxes and the examination rooms in the service area and operation room in the facility. Comparing with the other hot examination facilities in JAERI, WASTEF especially attends to different and particular works for investigating chemical behavior of waste forms. (author)

  8. 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

  9. Annual report on operation, utilization and technical Development of Hot Laboratories. From April 1, 1994 to March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-01-01

    This report describes activities, in fiscal year 1994, 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 rods irradiated in the Takahama Unit 3, a BWR fuel assembly irradiated in the Fukusima Daini Unit have been performed. Also, PIEs of ATR fuel assemblies and segment fuel assemblies 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 from the Halden reactor and TMI-2 debris samples. 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 have 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).

  10. Long-term high-level waste technology. Composite report

    Science.gov (United States)

    Cornman, W. R.

    1981-12-01

    Research and development studies on the immobilization of high-level wastes from the chemical reprocessing of nuclear reactor fuels are summarized. 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.

  11. 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.

  12. 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

  13. 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

  14. 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

  15. Gamma and X-ray shielding compositions utilizing bauxite - Red Mud regional research laboratory (CSIR), Bhopal, India

    International Nuclear Information System (INIS)

    Anshul, Avneesh; Amritphale, Sudhir Sitaram; Chandra, Navin; Ramakrishnan, N.

    2007-01-01

    Available in abstract form only. Full text of publication follows: The application spectrum of X-ray and Gamma radiation is increasing exponentially in the area of diagnostic, nuclear medicine, food preservation, nuclear power plants and strategic utilities. To prevent the harmful effects of these radiations, shielding materials based on lead metal and its compounds are being used historically, which are toxic in nature. To protect environment it has become necessary to develop non-toxic lead free shielding materials. The use of titanium metal and its compounds as synthetic rock i.e. SYNROC are reported to be very effective non-toxic shielding materials for various applications. Red mud waste generated in aluminum producing industries possesses a unique mineralogical compositions containing fairly high quantity of titanium oxide and iron oxide useful for making non toxic shielding compositions and therefore red mud has been utilized for the first time in the world for making radiation shielding materials. The red mud based compositions developed have been characterized for their various physico-mechanical properties namely compressive strength, impact strength, density and X-ray and gamma radiation shielding capacity in terms of shielding thickness i.e. HVT. Based on the characterization results it is found that the red mud based materials can be used for the construction of X-ray diagnostic and CT-Scanner room and as a substitute shielding material for concrete in the nuclear reactors and other radiation based applications. Studies on the identification of shielding phases and their morphology present, in the red mud based shielding compositions has been carried out using X-ray diffraction and SEM technique. The results of these studies are presented in this paper. (authors)

  16. Crichtonite structure type (AM21O38 and A2M19O36) as a host phase in crystalline waste form ceramics

    International Nuclear Information System (INIS)

    Gong, W.L.; Ewing, R.C.; Wang, L.M.; Xie, H.S.

    1995-01-01

    Previous studies of ceramic crystalline waste forms, e.g. Synroc, tailored ceramics, and supercalcine, have concentrated on phases which are major constituents in the formulations: zirconolite, pyrochlore, hollandite, perovskite and zircon. These phases usually occur as members of multi-phase assemblages which are required for the incorporation of the wide variety of radionuclide elements present in the waste and the non-radioactive components added during reprocessing and pretreatment. The crichtonite structure (AM 21 O 38 and A 2 M 19 O 36 ), based on crystallo-chemical considerations and natural compositional analogues, may effectively incorporate both fission products and actinides. The naturally occurring crichtonite structure types include Sr (crichtonite), Ca and REE (loveringite), Na (landauite), REE and U (davidite), K (mathiasite), Ba (lindsleyite), and Pb (senaite), which are classified based on the dominant, large cations occupying the A-site. The crystal structure contains three types of sites of distinct size, from very large, M 0 , intermediate (M 1 , M 3 , M 4 , and M 5 ), to small (M 2 ). Numerous coupled substitutions within these cation sites allow for charge balance. Synthesis experiments were completed on the Ba-, Sr-, Ca-, and K-member compositions at 3 GPa and 1,150 C. Low pressure synthesis should be possible, as natural minerals mostly occur in low-P systems. Reaction products were characterized by powder X-ray diffraction, scanning electron microscopy and electron microprobe analysis. In addition to the crichtonite phases, rutile, spinel, perovskite and armacolite were identified as well. The Crichtonite structure type is estimated to accommodate waste loading of up to 30 wt. % PW-4B waste

  17. 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

  18. 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)

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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(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 C and 1,100 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 Bi2Sr2Ca2Cu3O10 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 characterized by XRD

  1. 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

  2. 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

  3. R and D for actinide partitioning and recovery of valuables from high level waste using radiotracers

    International Nuclear Information System (INIS)

    Manchanda, V.K.

    2006-01-01

    In the context of growing world population with rapidly increasing energy needs and the threat of global warming due to CO 2 emission (caused by fossil fuel burning), the nuclear energy may be an attractive option particularly in the developing countries. Recycling of fuel is a unique feature of nuclear power technology which makes it a favourable choice with respect to conservation of energy resources. Steady growth of global fuel reprocessing activities (6000 tHM/annum) implies a vital role of separation science in developing efficient procedures for the separation and purification of actinides and in devising safe procedures for the management of nuclear waste arising at different stages of the PUREX process. High Level Waste (HLW) comprising of the concentrate of the raffinate of the co-extraction cycle (with over 95% of the total radioactivity produced in the burn up process in reactor) need to be isolated from the biosphere. There is a consensus among the waste management technologists that the safest route to achieve this, is to deposit it in a stable geological formation after it's immobilization in suitable glass/Synroc matrix. It ensures that any risk from exposure due to accidental intervention or natural disturbance is minimized. Risk perception is essentially due to the large radiological toxicity associated with alpha emitters like 237 Np, 241 Am, 243 Am and 245 Cm. Isotopes of Pu (left unrecovered) present in HLW also contribute towards radiological toxicity. In view of the high cost involved and the need for continuous surveillance, several countries are considering modifying their reprocessing schemes to partition (isolate) long-lived actinides from HLW. Since the volume of the actinide oxides (which retain major fraction of the radio toxicity of HLW) is significantly lower as compared to the other metal oxides present in HLW, such an approach is expected to reduce the cost of immobilization as well as of disposal (in geological repository) and

  4. Radioactive waste management at ANSTO - Managing current and historic wastes

    International Nuclear Information System (INIS)

    Harries, J.; Dimitrovski, L.; Hart, K.; Levins, D.

    2001-01-01

    developed involving concentration of the waste by evaporation, destruction of the ammonium ion by a novel process and solidification of the waste as a uranium-rich salt. Routine processing of the liquid waste commenced in 1999 and to date over 2 m 3 of liquid waste has been converted to a solid. The solidified waste is stored in high- integrity stainless steel vessels with a design life of at least 50 years. Another project under way will convert this solid waste into a more durable waste form suitable for long term storage or disposal. Two waste forms were initially considered; a titanate-based variant of synroc and cement. Laboratory scale testing established the feasibility of producing the titanate based ceramic with a high waste loading (∼44 wt % U) and the superior performance of this matrix over cement. Engineering scale development of a hot cell process for production of the ceramic waste form is under way. Much of the historic waste was characterised when it was generated by external dose with little information recorded about the radionuclide content. In 1996, a radioactive waste scanning system was installed to determine the radionuclide content of drums of historic waste. A data base system is being developed to integrate the characterisation, treatment and location information on the radioactive waste at ANSTO including the results from the drum scanning measurements. An important objective of ANSTO's waste management policy is minimization of radioactive waste generated and stored. This is being achieved by a number of strategies: for example, in one radioisotope production area a threefold reduction in waste volume has been achieved by separating non-radioactive waste from radioactive waste at the source. A substantial reduction in radioactive gas emissions during the production of molybdenum-99 has also been achieved by changes in waste processing operations and procedures. As well as focussing on historical waste issues a number of initiatives within

  5. 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