Performance Assessment of Zircon as a Waste Form for Excess Weapons Plutonium Under Deep Borehole Burial Conditions

1995 ◽  
Vol 412 ◽  
Author(s):  
W. J. Weber ◽  
R. C. Ewing ◽  
W. Lutze
Keyword(s):  
Performance Assessment ◽  
Radiation Damage ◽  
Alpha Decay ◽  
Chemical Durability ◽  
Waste Form ◽  
Accessory Mineral ◽  
Deep Borehole ◽  
Waste Forms ◽  
Weapons Grade Plutonium ◽  
Precambrian Terranes

AbstractZircon (ZrSiO4) is proposed as a waste form for excess weapons-grade plutonium. Zircon is an extremely durable ceramic that is often found as an accessory mineral in Precambrian terranes with ages up to 4 billion years. The chemical durability of zircon in groundwater far exceeds that of other waste forms, as modeled leach rates may be as low as 10-11g/m2d. At least 10 wt% Pu can substitute for Zr in zircon. Self-radiation damage from alpha decay leads to a crystalline-to-amorphous transformation that is modeled as a function of time and temperature for deep borehole conditions. Based on the results of this assessment, zircon could meet all necessary durability and criticality criteria required for a Pu waste form. The types of data used in this analysis are generally not available for other crystalline ceramics or glasses.

Ion Implantation Studies of Nuclear Waste Forms

1981 ◽  
Vol 6 ◽  
Author(s):  
Clyde J. M. Northrup ◽  
George W. Arnold ◽  
Thomas J. Headley
Keyword(s):  
Nuclear Waste ◽  
Borosilicate Glass ◽  
Alpha Decay ◽  
Lead Ions ◽  
Waste Form ◽  
Elastic Recoil ◽  
Elastic Recoil Detection ◽  
Waste Forms ◽  
Basic Studies ◽  
Nuclear Waste Forms

ABSTRACTThe first observations of physical and chemical changes induced by lead implantation damage and leaching are reported for two proposed U.S. nuclear waste forms (PNL 76–68 borosilicate glass and Sandia titanate ceramics) for commercial wastes. To simulate the effects of recoil nucleii due to alpha decay, the materials were implanted with lead ions at equivalent doses up to approximately 1 × 1019 a decays/cm3 . In the titanate waste form, the zirconolite, perovskite, hollandite, and rutile phases all exhibited a mottled appearance in the transmission electron microscope (TEM) typical of defect clusters in radiation damaged, crystalline solids. One titanate phase containing uranium was found by TEM to be amorphous after implantation at the highest dose. No enhanced leaching (deionized water, room temperature, 24 hours) of the irradiated titanate waste form, including the amorphous phase, was detected by TEM, but Rutherford backscattering (RBS) suggested a loss of cesium and calcium after 21 hours of leaching. The RBS spectra also indicated enhanced leaching from the PNL 76–68 borosilicate glass after implantation with lead ions, in general agreement with the observations of Dran, et al. [6,7] on other irradiated materials. Elastic recoil detection spectroscopy (ERD), used to profile hydrogen after leaching, showed penetration of the hydrogen to several thousand angstroms for both the implanted and unimplanted materials. These basic studies identified techniques to follow the changes that occur on implantation and leaching of complex amorphous and crystalline waste forms. These studies were not designed to produce comparisons between waste forms of gross leach rates.

Self-Irradiation of Ceramics and Single Crystals Doped With Pu-238: Summary of 5 Years of Research of the V. G. Khlopin Radium Institute

2008 ◽  
Vol 1107 ◽  
Author(s):  
Boris E. Burakov ◽  
Maria A. Yagovkina ◽  
Maria V. Zamoryanskaya ◽  
Vladimir M. Garbuzov ◽  
Vladimir A. Zirlin ◽  
...  
Keyword(s):  
Single Crystal ◽  
Radiation Damage ◽  
Single Crystals ◽  
Cumulative Dose ◽  
Crack Formation ◽  
Tetragonal Zirconia ◽  
Alpha Decay ◽  
Chemical Durability ◽  
Radium Institute ◽  
Cubic Zirconia

AbstractTo investigate the resistance of actinide host phases to accelerated radiation damage, which simulates radiation induced effects of long term storage, the following samples doped with plutonium-238 (from 2 to 10 wt. %) have been repeatedly studied using XRD and other methods: cubic zirconia, Zr0.79Gd0.14Pu0.07O1.99; monazite, (La,Pu)PO4; ceramic based on Pu-phosphate of monazite structure, PuPO4; ceramic based on zircon, (Zr,Pu)SiO4, and minor phase tetragonal zirconia, (Zr,Pu)O2; single crystal zircon, (Zr,Pu)SiO4; single crystal monazite, (Eu,Pu)PO4; ceramic based on Ti-pyrochlore, (Ca,Gd,Hf,Pu,U)2Ti2O7. No change of phase composition, matrix swelling, or cracking in cubic zirconia were observed after cumulative dose 2.77×1025 alpha decay/m3. The La-monazite remained crystalline at cumulative dose 1.19×1025 alpha decay/m3, although Pu-phosphate of monazite structure became nearly amorphous at relatively low dose 4.2×1024 alpha decay/m3. Zircon has lost crystalline structures under self-irradiation at dose (1.3-1.5)×1025 alpha decay/m3, however, amorphous zircon characterized with high chemical durability. The Ti-pyrochlore after cumulative dose (1.1-1.3)×1025 alpha decay/m3 became amorphous and lost chemical durability. Radiation damage caused crack formation in zircon single crystals but not in the matrix of polycrystalline zircon. Essential swelling and crack formation as a result of radiation damage were observed in ceramics based on Ti-pyrochlore and Pu-phosphate of monazite structure, but not so far in La-monazite doped with 238Pu.

scholarly journals Post-Closure Performance Assessment for Deep Borehole Disposal of Cs/Sr Capsules

Energies ◽  
2019 ◽  
Vol 12 (10) ◽  
pp. 1980 ◽  
Author(s):  
Geoff Freeze ◽  
Emily Stein ◽  
Patrick Brady
Keyword(s):  
Performance Assessment ◽  
Waste Disposal ◽  
Host Rock ◽  
Department Of Energy ◽  
Waste Form ◽  
Deep Borehole ◽  
Waste Package ◽  
No Releases

Post-closure performance assessment (PA) calculations suggest that deep borehole disposal of cesium (Cs)/strontium (Sr) capsules, a U.S. Department of Energy (DOE) waste form (WF), is safe, resulting in no releases to the biosphere over 10,000,000 years when the waste is placed in a 3–5 km deep waste disposal zone. The same is true when a hypothetical breach of a stuck waste package (WP) is assumed to occur at much shallower depths penetrated by through-going fractures. Cs and Sr retardation in the host rock is a key control over movement. Calculated borehole performance would be even stronger if credit was taken for the presence of the WP.

Stability of Cubic Zirconia in a Granitic System Under High Pressure & Temperature

2008 ◽  
Vol 1107 ◽  
Author(s):  
Fergus G.F. Gibb ◽  
Boris E. Burakov ◽  
Kathleen J. Taylor ◽  
Yana Domracheva
Keyword(s):  
Radioactive Wastes ◽  
Waste Form ◽  
Deep Borehole ◽  
Spent Fuel ◽  
Zirconia Crystal ◽  
Cubic Zirconia ◽  
Inert Matrix ◽  
Waste Forms ◽  
Glassy Sample ◽  
Container Material

AbstractCubic zirconia is a well known, highly durable material with potential uses as an actinide host phase in ceramic waste forms and inert matrix fuels and in containers for very deep borehole disposal of some highly radioactive wastes. To investigate the behaviour of this material under the conditions of possible use, a cube of ∼ 2.5 mm edge was made from a single crystal of yttriastabilized cubic zirconia doped with 0.3 wt.% CeO2. The cube was enclosed in powdered granite within a gold capsule and a small amount of H2O added before sealing. The sealed capsule was held for 4 months in a cold-seal pressure vessel at a temperature of 780°C and a pressure 150 MPa, simulating both the conditions of a deep borehole disposal involving partial melting of the host rock and the conditions under which the actinide waste form might be encapsulated in granite prior to disposal. At the end of the experiment the quenched, largely glassy, sample was cut into thin slices and studied by optical microscopy, EMPA, SEM and cathodoluminescence methods. The results show that no corrosion of the zirconia crystal or reaction with the granite melt occurred and that no detectable diffusion of elements, including Ce, in or out of the zirconia took place on the timescale of the experiment. Consequently, it appears that cubic zirconia could perform most satisfactorily as both an actinide host waste form for encapsulation in solid granite for very deep disposal and as a container material for deep borehole disposal of highly radioactive wastes (HLW), including spent fuel.

Comparison of Glass and Crystalline Nuclear Waste Forms

1988 ◽  
Vol 127 ◽  
Author(s):  
W. Lutze ◽  
R. C. Ewing
Keyword(s):  
Mechanical Properties ◽  
Performance Assessment ◽  
Nuclear Waste ◽  
Single Phase ◽  
Chemical Durability ◽  
Borosilicate Glasses ◽  
Thermal And Mechanical Properties ◽  
Materials Properties ◽  
Waste Forms ◽  
Nuclear Waste Forms

ABSTRACTNuclear waste forms may be divided into two broad categories: (1) single phase glasses with minor crystalline components (e.g., borosilicate glasses) and (2) crystalline waste forms, either single phase (e.g., monazite) or polyphase (e.g., SYNROC). This paper reviews the materials properties data that are available for each of these two types of waste forms. The prinicipal data include: (1) physical, thermal and mechanical properties, (2) chemical durability; (3) radiation damage effects. Complete data are only available for borosilicate glasses and SYNROC; therefore, this comparison focuses on the performance assessment of borosilicate glass and SYNROC.

scholarly journals Preparation and characterization of multiphase ceramic designer waste forms

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Braeden M. Clark ◽  
Priyatham Tumurugoti ◽  
Shanmugavelayutham K. Sundaram ◽  
Jake W. Amoroso ◽  
James C. Marra
Keyword(s):  
Fixed Ratio ◽  
Chemical Durability ◽  
Waste Form ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Plasma Sintering ◽  
Long Term Performance ◽  
Term Performance ◽  
Elemental Release

AbstractThe long-term performance, or resistance to elemental release, is the defining characteristic of a nuclear waste form. In the case of multiphase ceramic waste forms, correlating the long-term performance of multiphase ceramic waste forms in the environment to accelerated chemical durability testing in the laboratory is non-trivial owing to their complex microstructures. The fabrication method, which in turn affects the microstructure, is further compounding when comparing multiphase ceramic waste forms. In this work, we propose a “designer waste form” prepared via spark plasma sintering to limit interaction between phases and grain growth during consolidation, leading to monolithic high-density waste forms, which can be used as reference materials for comparing the chemical durability of multiphase waste forms. Designer waste forms containing varying amounts of hollandite in the presence of zirconolite and pyrochlore in a fixed ratio were synthesized. The product consistency test (PCT) and vapor hydration test (VHT) were used to assess the leaching behavior. Samples were unaffected by the VHT after 1500 h. As measured by the PCT, the fractional Cs release decreased as the amount of hollandite increased. Elemental release from the zirconolite and pyrochlore phases did not appear to significantly contribute to the elemental release from the hollandite phase in the designer waste forms.

scholarly journals Intergrowth Structures in Synthetic Pyrochlores: Implications for Radiation Damage Effects and Waste Form Formulation

1999 ◽  
Vol 556 ◽  
Author(s):  
E. C. Buck ◽  
D. B. Chamberlain ◽  
R. Gieré
Keyword(s):  
Radiation Damage ◽  
Volume Expansion ◽  
Stress Relief ◽  
Waste Form ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Defense In Depth ◽  
Intergrowth Structures

AbstractTitanate-based ceramic waste forms are currently under development for the immobilization of excess weapons plutonium. Both Hf and Gd are added to the ceramic formulation as neutron absorbers in order to satisfy a defense-in-depth concept for the waste form. The introduction of significant amounts of hafnium may be responsible for the presence of zirconolite-2M crystals in pyrochlore-based ceramics and the formation of zirconolite lamellae within pyrochlore. The zirconolite grows epitaxially on { 111 }planes of pyrochlore. Although the zirconolite lamellae within pyrochlore are non-cubic, any volume expansion due to radiation damage in the pyrochlore should still be isotropic; in addition, the presence of these intergrowths may allow some stress relief in the ceramic.

Host Phases for Actinide Elements in the Metallic Waste Form

2002 ◽  
Vol 757 ◽  
Author(s):  
D. E. Janney
Keyword(s):  
Stainless Steel ◽  
Electron Microscope ◽  
Electron Diffraction ◽  
Transmission Electron Microscope ◽  
Waste Form ◽  
Dissolution Behavior ◽  
Release Rates ◽  
Waste Forms ◽  
Transmission Electron ◽  
Simulated Waste

ABSTRACTArgonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists of stainless steel cladding hulls that contain undissolved metal fission products such as Tc, Ru, Rh, Pd, and Ag; a small amount of undissolved actinides (U, Np, Pu) also remains with the hulls. These wastes will be immobilized in a waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). Scanning electron microscope (SEM) observations of simulated metal waste forms (SS-15Zr with up to 11 wt% actinides) show eutectic intergrowths of Fe-Zr-Cr-Ni intermetallic phases with steels. The actinide elements are almost entirely in the intermetallics, where they occur in concentrations ranging from 1–20 at%. Neutron- and electron-diffraction studies of the simulated waste forms show materials with structures similar to those of Fe2Zr and Fe23Zr6.Dissolution experiments on simulated waste forms show that normalized release rates of U, Np, and Pu differ from each other and from release rates of other elements in the sample, and that release rates for U exceed those for any other element (including Fe). This paper uses transmission electron microscope (TEM) observations and results from energy-dispersive X-ray spectroscopy (EDX) and selected-area electron-diffraction (SAED) to characterize relationships between structural and chemical data and understand possible reasons for the observed dissolution behavior.Transmission electron microscope observations of simulated waste form samples with compositions SS-15Zr-2Np, SS-15Zr-5U, SS-15Zr-11U-0.6Rh-0.3Tc-0.2Pd, and SS-15Zr-10Pu suggest that the major actinide-bearing phase in all of the samples has a structure similar to that of the C15 (cubic, MgCu2-type) polymorph of Fe2Zr, and that materials with this structure exhibit significant variability in chemical compositions. Material whose structure is similar to that of the C36 (dihexagonal, MgNi2-type) polymorph of Fe2Zr was also observed, and it exhibits less chemical variability than that displayed by material with the C15 structure. The TEM data also demonstrate a range of actinide concentrations in materials with the Fe23Zr6 (cubic, Mn23Th6-type) structure.Microstructures similar to those produced during experimental deformation of Fe-10 at% Zr alloys were observed in intermetallic materials in all of the simulated waste form samples. Stacking faults and associated dislocations are common in samples with U, but rarely observed in those with Np and Pu, while twins occurred in all samples. The observed differences in dissolution behavior between samples with different actinides may be related to increased defect-assisted dissolution in samples with U.

Aging Studies of Pu-238 and -239 Containing Calcium Phosphate Ceramic Waste-forms

2013 ◽  
Vol 1518 ◽  
pp. 73-78 ◽  
Author(s):  
Shirley K. Fong ◽  
Brian L. Metcalfe ◽  
Randall D. Scheele ◽  
Denis M. Strachan
Keyword(s):  
Calcium Phosphate ◽  
Accelerated Aging ◽  
Waste Form ◽  
Calcium Phosphate Ceramic ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Aging Studies ◽  
Flow Through ◽  
Pyrochemical Reprocessing

ABSTRACTA calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238, these were characterised by PXRD at regular intervals and Single Pass Flow Through (SPFT) tests after approximately 5 yrs.While XRD indicated some loss of crystallinity in the Pu-238 samples after exposure to 2.8 x 1018 α decays, SPFT tests indicated that accelerated aging had not had a detrimental effect on the durability of Pu-238 samples compared to Pu-239 waste-forms.

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