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.

TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy

2000 ◽  
Vol 6 (S2) ◽  
pp. 368-369
Author(s):  
N.L. Dietz ◽  
D.D Keiser
Keyword(s):  
Stainless Steel ◽  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Metallic Sodium ◽  
Geologic Repository ◽  
Tem Analysis ◽  
Waste Forms

Argonne National Laboratory has developed an electrometallurgical treatment process for metallic spent nuclear fuel from the Experimental Breeder Reactor-II. This process stabilizes metallic sodium and separates usable uranium from fission products and transuranic elements that are contained in the fuel. The fission products and other waste constituents are placed into two waste forms: a ceramic waste form that contains the transuranic elements and active fission products such as Cs, Sr, I and the rare earth elements, and a metal alloy waste form composed primarily of stainless steel (SS), from claddings hulls and reactor hardware, and ∼15 wt.% Zr (from the U-Zr and U-Pu-Zr alloy fuels). The metal waste form (MWF) also contains noble metal fission products (Tc, Nb, Ru, Rh, Te, Ag, Pd, Mo) and minor amounts of actinides. Both waste forms are intended for eventual disposal in a geologic repository.

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.

Performance of Borosilicate Glass High-Level Waste Forms in Disposal Systems

1986 ◽  
Vol 73 (2) ◽  
pp. 139-139
Author(s):  
Edward J. Hennelly ◽  
E. I. Du Pont de Nemours
Keyword(s):  
Borosilicate Glass ◽  
High Level Waste ◽  
Waste Forms ◽  
High Level

Fluidized Bed Steam Reformed (FBSR) Mineral Waste Forms: Characterization and Durability Testing

2006 ◽  
Vol 985 ◽  
Author(s):  
Carol Jantzen ◽  
Troy H. Lorier ◽  
John M. Pareizs ◽  
James C. Marra
Keyword(s):  
Fluidized Bed ◽  
Environmental Protection Agency ◽  
National Laboratory ◽  
Pilot Scale ◽  
Waste Form ◽  
High Sodium ◽  
Waste Forms ◽  
Ring Structures ◽  
Low Activity ◽  
Mineral Waste

AbstractFluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium low activity wastes (LAW) such as those existing at the Hanford site, at the Idaho National Laboratory (INL), and the Savannah River Site (SRS). The addition of clay, charcoal, and a catalyst as co-reactants with the waste denitrates the aqueous wastes and forms a granular mineral waste form that can subsequently be made into a monolith for disposal if necessary. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage and ring structures and iron bearing spinel minerals. The mineralization occurs at moderate temperatures between 650-750°C in the presence of superheated steam. The cage and ring structured feldspathoid minerals atomically bond radionuclides like Tc-99 and Cs-137 and anions such as SO4, I, F, and Cl. The spinel minerals stabilize Resource Conservation and Recovery Act (RCRA) hazardous species such as Cr and Ni. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW) in pilot scale facilities at the Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The results of the SPFT testing and the activation energies for dissolution are discussed in this study.

scholarly journals Development of a ceramic waste form for high-level waste disposal

1999 ◽  
Vol 556 ◽  
Author(s):  
D. W. Esh ◽  
K. M. Goff ◽  
K. T. Hirsche ◽  
T. J. Battisti ◽  
M. F. Simpson ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
High Level Waste ◽  
Durability Test ◽  
Ceramic Waste ◽  
High Level ◽  
The Impact

AbstractA ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.

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.

New Crystalline Silicotitanate (CST) Waste Forms: Hydrothermal Synthesis and Characterization of CS-SI-TI-O Phases

1999 ◽  
Vol 556 ◽  
Author(s):  
M. Nyman ◽  
T. M. Nenoff ◽  
Y. Su ◽  
M. L. Balmer ◽  
A. Navrotsky ◽  
...  
Keyword(s):  
Thermal Stability ◽  
Hydrothermal Synthesis ◽  
Ion Exchange ◽  
Borosilicate Glass ◽  
Waste Form ◽  
Synthesis And Characterization ◽  
Hanford Site ◽  
Waste Forms ◽  
Crystalline Silicotitanate

AbstractThe radioactivity of the Hanford site waste tanks is primarily from 137Cs and 90Sr, of which can both be selectively removed from solution using a crystalline silicotitanate (CST) ion exchanger. We are currently seeking waste forms alternative to borosilicate glass for Cs-CSTs. In order to obtain a fundamental basis for the development of an alternative waste form, we are investigating synthesis and characterization of CST component phases, namely Cs-Si-Ti-O phases. Two novel Cs-Ti-Si-O phases (one porous, one condensed) have been hydrothermally synthesized, characterized and evaluated as waste form candidates based on chemical and thermal stability, leachability, and ion exchange capabilities.

scholarly journals Selection of Barrier Metals for a Waste Package in Tuff

1983 ◽  
Vol 26 ◽  
Author(s):  
E. W. yyRussell ◽  
R. D. McCright ◽  
W. C. O'Neal
Keyword(s):  
Nuclear Waste ◽  
Lawrence Livermore National Laboratory ◽  
Near Field ◽  
National Laboratory ◽  
Test Site ◽  
Management Program ◽  
Nuclear Waste Storage ◽  
Waste Storage ◽  
High Level ◽  
Selection Of

ABSTRACTThe Nevada Nuclear Waste Storage Investigations (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. Lawrence Livermore National Laboratory is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. In this paper we address the selection of metal containment barriers.

Evaluation of Cesium Sujcotitanates as an Alternative Waste Form

1996 ◽  
Vol 465 ◽  
Author(s):  
Yali Su ◽  
M. Lou Balmer ◽  
Bruce C. Bunker
Keyword(s):  
Borosilicate Glass ◽  
Thermal Conversion ◽  
Chemical Durability ◽  
Waste Form ◽  
High Level Waste ◽  
Ion Exchangers ◽  
Heat Treated ◽  
High Level ◽  
Excellent Chemical ◽  
Better Than

ABSTRACTSilicotitanate ion exchangers are potential materials for the removal of radioactive Cs and Sr from tank wastes. In this paper the viability of direct thermal conversion of Cs-loaded silicotitanates to an acceptable high level waste form has been examined. Results show that in aqueous solutions, the Cs leach rates of crystalline silicotitanates (heat treated at 800°C) are 0.04, 0.18, 0.4 g/m2day for Cs loadings of 1, 5, and 20 wt%, respectively. Heating the Cs-loaded (up to 20 wt %) silicotitanates at or above 900 °C for 1 hour further reduces the Cs leach rates to approximately zero (beyond the lppm detection limits). Moreover, Cs volatilization was found to be < 0.8 wt% at temperatures as high as 1000 °C. These results suggest that thermally converted silicotitanate ion exchangers exhibit excellent chemical durability (comparable to or better than borosilicate glass) and thus, have great potential as an alternative waste form.

An Experimental Comparison of Alternative Solid Forms for Savannah River High-Level Wastes

1981 ◽  
Vol 6 ◽  
Author(s):  
John A. Stone
Keyword(s):  
Borosilicate Glass ◽  
Silica Glass ◽  
Deionized Water ◽  
Experimental Comparison ◽  
Savannah River ◽  
Waste Forms ◽  
High Silica ◽  
High Level ◽  
Selection Of ◽  
High Aluminum

ABSTRACTSamples of borosilicate glass, high-silica glass, tailored ceramic, and SYNROC, incorporating simulated Savannah River high-level defense waste sludges, were leached by the MCC-1 procedure for times up to 28 days. Cesium, uranium, and cerium leach rates are reported for waste forms containing a composite sludge, at 40°C in deionized water, and at 90°C in deionized water, silicate water, and brine. The ordering of the waste forms from best to worst differs for each element leached, and none of the forms show a clear advantage for all the key radwaste elements. Some cesium leach rates for forms containing high-aluminum or high-iron sludges also are presented. So far, only small effects of sludge type have been observed, with one exception. This study is one of several inputs for selection of an alternative waste form for Savannah River waste.

Sign in / Sign up

Export Citation Format

Share Document