Interim Waste Forms for High-Level Radioactive Wastes: Processing and Properties

1981 ◽  
Vol 6 ◽  
Author(s):  
G. Bandyopadhyay
Keyword(s):  
Sodium Silicate ◽  
Water Solubility ◽  
Mechanical Stability ◽  
Solid Material ◽  
High Water ◽  
Radioactive Wastes ◽  
Waste Form ◽  
Waste Forms ◽  
High Level ◽  
Fused Salt

ABSTRACTSeveral simulated interim waste forms have been investigated in the laboratory to study their suitability for application in handling and transportation of high-level radioactive wastes to terminal processing sites. In the fused-salt/sludge option, the neutralized supernatant liquid and the precipitated sludge are treated simultaneously to form fused-salt cakes. Silicate-based options, in which sodium silicate or sodium silicate and Ca(OH)2 act as binders for the sludge, require prior separation of the sludge and the soluble radioactive constituents from the supernatant before the waste form can be prepared. The results from tests on simulated fused-salt waste forms indicated that the process simplicity of this option is partially offset by the high water solubility and hygroscopicity of the product, which would necessitate special precautions during transportation and storage. The most promising silicate-based option is the ambienttemperature silicate sludge process, in which the sludge is mixed with sodium silicate [and sometimes with Ca(OH)2] and subsequently exposed to a contrelled-humidity environment at room temperature to form a chemical bond. Solid material containing 75 wt % synthetic calcined sludge, prepared by this process, has sufficient physical, chemical, and mechanical stability for use as an interim waste form.

scholarly journals Incorporation of High-Level Wastes in Synroc: Results from Recent Process Engineering Studies at Lawrence Livermore National Laboratory

1981 ◽  
Vol 11 ◽  
Author(s):  
J. H. Campbell ◽  
C. L. Hoenig ◽  
F. J. Ackerman ◽  
P. E. Peters ◽  
J. Z. Grens
Keyword(s):  
Borosilicate Glass ◽  
Selection Process ◽  
Lawrence Livermore National Laboratory ◽  
National Laboratory ◽  
Process Engineering ◽  
Waste Form ◽  
Performance Properties ◽  
Engineering Studies ◽  
Waste Forms ◽  
High Level

In October 1981 SYNROC-D was selected as the reference alternate waste form to borosilicate glass for immobilization of defense wastes. A total of eight candidate waste forms competed in this selection process and the decision of which alternate waste form to choose was based primarily on performance properties.

The Strategy of the Belgian Nuclear Research Centre in the Area of High-Level Waste Form Compatibility Research

2007 ◽  
Author(s):  
Karel Lemmens ◽  
Christelle Cachoir ◽  
Elie Valcke ◽  
Karine Ferrand ◽  
Marc Aertsens ◽  
...  
Keyword(s):  
Nuclear Research ◽  
Waste Form ◽  
Research Centre ◽  
High Level Waste ◽  
Spent Fuel ◽  
National Agency ◽  
Waste Forms ◽  
In Situ Experiments ◽  
High Level

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.

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.

Corrosion Testing of Stainless Steel-Zirconium:Metal Waste Forms

1999 ◽  
Vol 556 ◽  
Author(s):  
D. P. Abraham ◽  
L. J. Simpson ◽  
M. J. Devries ◽  
S. M. Mcdeavitt
Keyword(s):  
Stainless Steel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Waste Form ◽  
Waste Forms ◽  
Immersion Corrosion ◽  
Corrosion Studies ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Nuclear Waste Forms

AbstractStainless steel-zirconium (SS-Zr) alloys have been developed as waste forms for the disposal of metallic waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste forms incorporate irradiated cladding hulls, components of the alloy fuel, noble metal fission products, and actinide elements. The baseline waste form is a stainless steel- 15 wt% zirconium (SS-15Zr) alloy. This article presents microstructures and some of the corrosion studies being conducted on the waste form alloys. Electrochemical corrosio n, immersion corrosion, and vapor hydration tests have been performed on various alloy compositions to evaluate corrosion behavior and resistance to selective leaching of simulated fission products. The SS-Zr waste forms immobilize and retain fission products very effectively and show potential for acceptance as high-level nuclear waste forms.

Application of the TSPA Glass Degradation Model to Non-Conforming Waste Forms

2004 ◽  
Vol 824 ◽  
Author(s):  
W. L. Ebert
Keyword(s):  
Waste Form ◽  
Total System ◽  
Borosilicate Glasses ◽  
Degradation Mechanisms ◽  
Release Rates ◽  
Degradation Model ◽  
Waste Forms ◽  
Glass Model ◽  
High Level ◽  
Fractional Release

AbstractAn approach is presented for determining if the models used to calculate the release of radionuclides from defense high-level radioactive waste (HLW) glass for total system performance assessment (TSPA) calculations can be used to account for the release of radionuclides from waste forms other than standard borosilicate glasses. The fractional release rates of radionuclides due to waste form degradation, the available surface area, and the radionuclide inventory in an alternative waste form can be compared with the corresponding models used in TSPA for HLW glasses to determine if those models adequately represent the waste form. This approach is demonstrated for the ceramic and metallic waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel. Depending on the waste form, comparisons made with aspects of the HLW glass model may be based on similarities in degradation mechanisms or purely empirical.

Sodium Zirconium Phosphate [NZP] as a Host Matrix for High Level Radioactive Wastes

2002 ◽  
Vol 757 ◽  
Author(s):  
Yoshimi Seida ◽  
Mami Yuki ◽  
Kazunori Suzuki ◽  
Toshio Sawa
Keyword(s):  
Zirconium Phosphate ◽  
Sol Gel ◽  
Radioactive Wastes ◽  
Charge Balance ◽  
Small Disk ◽  
Host Matrix ◽  
Waste Forms ◽  
Sodium Zirconium ◽  
High Level ◽  
Sodium Zirconium Phosphate

ABSTRACTVarious elements (Cs, Sr, Ba, Zr, Ru, Pd, Ce, Nd, Gd, Fe, Cr, Ni, Mo and Te) in the simulated high-level radioactive wastes generated from commercial PUREX reprocessing were immobilized by a ceramic solidification using sodium zirconium phosphate, NaZr2(PO4)3 as a host matrix. The convertibility of the elements to the specific M, A and X sites in NZP crystal structure was determined with consideration of stoichiometry, charge balance and ion size of each element. Small disk samples of NZP waste form containing the elements were prepared by the sol-gel synthesis followed by calcination and compression sinteration at high temperature. The physicochemical structures such as produced phase in the waste forms and dispersion of the embedded elements in the NZP waste forms were investigated by means of XRD and SEM/EDX. Chemical behavior of the embedded elements and the limit of substitution of the NZP structure for the elements were investigated by the series of analysis. Moreover, in order to increase the limit of substitution of NZP, the effectiveness of pretreatment with heating the sol-gel products at 473–773K was experimentally investigated. The capability of embedding of NZP for the elements in HLW has been discussed.

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