24 Matching Results

Search Results

Advanced search parameters have been applied.

National high-level waste systems analysis

Description: Previously, no mechanism existed that provided a systematic, interrelated view or national perspective of all high-level waste treatment and storage systems that the US Department of Energy manages. The impacts of budgetary constraints and repository availability on storage and treatment must be assessed against existing and pending negotiated milestones for their impact on the overall HLW system. This assessment can give DOE a complex-wide view of the availability of waste treatment and help project the time required to prepare HLW for disposal. Facilities, throughputs, schedules, and milestones were modeled to ascertain the treatment and storage systems resource requirements at the Hanford Site, Savannah River Site, Idaho National Engineering Laboratory, and West Valley Demonstration Project. The impacts of various treatment system availabilities on schedule and throughput were compared to repository readiness to determine the prudent application of resources. To assess the various impacts, the model was exercised against a number of plausible scenarios as discussed in this paper.
Date: May 1, 1996
Creator: Kristofferson, K. & O`Holleran, T.P.
Partner: UNT Libraries Government Documents Department

Crystalline plutonium hosts derived from high-level waste formulations.

Description: The Department of Energy has selected immobilization for disposal in a repository as one approach for disposing of excess plutonium (1). Materials for immobilizing weapons-grade plutonium for repository disposal must meet the ''spent fuel standard'' by providing a radiation field similar to spent fuel (2). Such a radiation field can be provided by incorporating fission products from high-level waste into the waste form. Experiments were performed to evaluate the feasibility of incorporating high-level waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) into plutonium dispositioning materials to meet the spent fuel standard. A variety of materials and preparation techniques were evaluated based on prior experience developing waste forms for immobilizing HLW. These included crystalline ceramic compositions prepared by conventional sintering and hot isostatic pressing (HIP), and glass formulations prepared by conventional melting. Because plutonium solubility in silicate melts is limited, glass formulations were intentionally devitrified to partition plutonium into crystalline host phases, thereby allowing increased overall plutonium loading. Samarium, added as a representative rare earth neutron absorber, also tended to partition into the plutonium host phases. Because the crystalline plutonium host phases are chemically more inert, the plutonium is more effectively isolated from the environment, and its attractiveness for proliferation is reduced. In the initial phase of evaluating each material and preparation method, cerium was used as a surrogate for plutonium. For promising materials, additional preparation experiments were performed using plutonium to verify the behavior of cerium as a surrogate. These experiments demonstrated that cerium performed well as a surrogate for plutonium. For the most part, cerium and plutonium partitioned onto the same crystalline phases, and no anomalous changes in oxidation state were observed. The only observed difference in behavior between cerium and plutonium was that plutonium partitioned more completely into the major host phases than cerium. Where cerium ...
Date: April 24, 1998
Creator: O'Holleran, T. P.
Partner: UNT Libraries Government Documents Department

A Versatile two-step process for immobilizing excess plutonium.

Description: As a consequence of weapon stockpile reduction and the associated shutdown of weapons production facilities, approximately 50 metric tons of plutonium (both weapons-grade and non-weapons-grade) has been declared excess by the US. Recent experiments demonstrated the feasibility of using high-level waste stored at the Idaho Chemical Processing Plant to immobilize plutonium. The most effective plutonium host phase identified in these experiments was a plutonium zirconate solid solution. Results of recent experiments are reported that show the feasibility of using the highly durable plutonium zirconate host phase as a feed material for high and low temperature encapsulation processes, thereby increasing the potential applications of this material for plutonium dispositioning.
Date: May 18, 1998
Creator: O'Holleran, T. P.
Partner: UNT Libraries Government Documents Department

Development of a sampling method for qualification of a ceramic high-level waste form.

Description: A ceramic waste form has been developed to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The ceramic waste form was originally prepared in a hot isostatic press (HIP). Small HIP capsules called witness tubes were used to obtain representative samples of material for process monitoring, waste form qualification, and archiving. Since installation of a full-scale HIP in existing facilities proved impractical, a new fabrication process was developed. This process fabricates waste forms inside a stainless steel container using a conventional furnace. Progress in developing a new method of obtaining representative samples is reported.
Date: July 2, 2002
Creator: O'Holleran, T. P.
Partner: UNT Libraries Government Documents Department

Microwave heating for production of a glass bonded ceramic high-level waste form.

Description: Argonne National Laboratory has developed a ceramic waste form to immobilize the salt waste from electrometallurgical treatment of spent nuclear fuel. The process is being scaled up to produce bodies of 100 Kg or greater. With conventional heating, heat transfer through the starting powder mixture necessitates long process times. Coupling of 2.45 GHz radiation to the starting powders has been demonstrated. The radiation couples most strongly to the salt occluded zeolite powder. The results of these experiments suggest that this ceramic waste form could be produced using microwave heating alone, or by using microwave heating to augment conventional heating.
Date: July 30, 2002
Creator: O'Holleran, T. P.
Partner: UNT Libraries Government Documents Department

Preliminary Waste Form Compliance Plan for the Idaho National Engineering and Environmental Laboratory High-Level Waste

Description: The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.
Date: May 1, 1999
Creator: Staples, B. A. & O'Holleran, T. P.
Partner: UNT Libraries Government Documents Department

Comparison of mechanical properties of glass-bonded sodalite and borosilicate glass high-level waste forms

Description: Argonne National Laboratory has developed a glass-bonded sodalite waste form to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The waste form consists of 75 vol.% crystalline sodalite and 25 vol.% glass. Microindentation fracture toughness measurements were performed on this material and borosilicate glass from the Defense Waste Processing Facility using a Vickers indenter. Palmqvist cracking was confined for the glass-bonded sodalite waste form, while median-radial cracking occurred in the borosilicate glass. The elastic modulus was measured by an acoustic technique. Fracture toughness, microhardness, and elastic modulus values are reported for both waste forms.
Date: May 9, 2000
Creator: O'Holleran, T. P.; DiSanto, T.; Johnson, S. G. & Goff, K. M.
Partner: UNT Libraries Government Documents Department

Fracture toughness measurements on a glass bonded sodalite high-level waste form.

Description: The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies.
Date: May 19, 1999
Creator: DiSanto, T.; Goff, K. M.; Johnson, S. G. & O'Holleran, T. P.
Partner: UNT Libraries Government Documents Department

Reference Alloy Waste Form Fabrication and Initiation of Reducing Atmosphere and Reductive Additives Study on Alloy Waste Form Fabrication

Description: This report describes the fabrication of two reference alloy waste forms, RAW-1(Re) and RAW-(Tc) using an optimized loading and heating method. The composition of the alloy materials was based on a generalized formulation to process various proposed feed streams resulting from the processing of used fuel. Waste elements are introduced into molten steel during alloy fabrication and, upon solidification, become incorporated into durable iron-based intermetallic phases of the alloy waste form. The first alloy ingot contained surrogate (non-radioactive), transition-metal fission products with rhenium acting as a surrogate for technetium. The second alloy ingot contained the same components as the first ingot, but included radioactive Tc-99 instead of rhenium. Understanding technetium behavior in the waste form is of particular importance due the longevity of Tc-99 and its mobility in the biosphere in the oxide form. RAW-1(Re) and RAW-1(Tc) are currently being used as test specimens in the comprehensive testing program investigating the corrosion and radionuclide release mechanisms of the representative alloy waste form. Also described in this report is the experimental plan to study the effects of reducing atmospheres and reducing additives to the alloy material during fabrication in an attempt to maximize the oxide content of waste streams that can be accommodated in the alloy waste form. Activities described in the experimental plan will be performed in FY12. The first aspect of the experimental plan is to study oxide formation on the alloy by introducing O2 impurities in the melt cover gas or from added oxide impurities in the feed materials. Reducing atmospheres will then be introduced to the melt cover gas in an attempt to minimize oxide formation during alloy fabrication. The second phase of the experimental plan is to investigate melting parameters associated with alloy fabrication to allow the separation of slag and alloy components of the melt.
Date: September 1, 2011
Creator: Frank, S.M.; O'Holleran, T.P. & Hahn, P.A.
Partner: UNT Libraries Government Documents Department

Preparation of plutonium waste forms with ICPP calcined high-level waste

Description: Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.
Date: May 1, 1997
Creator: Staples, B.A.; Knecht, D.A. & O`Holleran, T.P.
Partner: UNT Libraries Government Documents Department

Analytical electron microscopy study of radioactive ceramic waste form

Description: A ceramic waste form has been developed to immobilize the halide high-level waste stream from electrometallurgical treatment of spent nuclear fuel. Analytical electron microscopy studies, using both scanning and transmission instruments, have been performed to characterize the microstructure of this material. The microstructure consists primarily of sodalite granules (containing the bulk of the halides) bonded together with glass. The results of these studies are discussed in detail. Insight into the waste form fabrication process developed as a result of these studies is also discussed.
Date: November 11, 1999
Creator: O'Holleran, T. P.; Sinkler, W.; Moschetti, T. L.; Johnson, S. G. & Goff, K. M.
Partner: UNT Libraries Government Documents Department

Characterization of a glass-bonded ceramic waste form loaded with U and Pu

Description: This paper presents microscopic characterization of four samples of a ceramic waste form (CWF) developed for disposal of actinide-containing electrorefiner salts. The four samples were prepared to investigate the influence of water content and the Pu:U ratio on CWF microstructure and performance. While the overall phase content is not strongly influenced by either variable, the presence of water in the initial zeolite has a detectable effect on CWF microstructure. It is found to influence the distribution of the major actinide host phase, a (U,Pu)O{sub 2} mixed oxide.
Date: November 19, 1999
Creator: Sinkler, W.; O'Holleran, T. P.; Frank, S. M.; Richmann, M. K. & Johnson, S. G.
Partner: UNT Libraries Government Documents Department

Microstructure and phase formation in a 17 weight percent plutonium oxide devitrified waste glass

Description: Plutonium containing ceramic waste forms have been prepared by dry mixing and melting glass frit, simulated zirconia high level waste calcine from chemical reprocessing, samarium oxide, titanium metal, and plutonium oxide. Materials were produced using melt times of 4 and 12 hours at 1,450 C followed by a thermal anneal at 500 C. Complex materials with a substantial volume fraction of crystalline phases were the result. The principle plutonium bearing phase was identified as a fluorite structured plutonium-zirconium-samarium phase of variable stoichiometry. This high plutonium phase preferentially segregated to the bottom of the waste form. A waste form was also melted using metallic plutonium in a quantity equivalent to 15 wt% plutonium oxide. XRD results indicate that the metal was completely oxidized on melting.
Date: September 1, 1997
Creator: Meyer, M.K.; Johnson, S.G.; O`Holleran, T.P. & Frank, S.M.
Partner: UNT Libraries Government Documents Department

Characterization of composite ceramic high level waste forms.

Description: Argonne National Laboratory has developed a composite ceramic waste form for the disposition of high level radioactive waste produced during electrometallurgical conditioning of spent nuclear fuel. The electrorefiner LiCl/KCl eutectic salt, containing fission products and transuranics in the chloride form, is contacted with a zeolite material which removes the fission products from the salt. After salt contact, the zeolite is mixed with a glass binder. The zeolite/glass mixture is then hot isostatic pressed (HIPed) to produce the composite ceramic waste form. The ceramic waste form provides a durable medium that is well suited to incorporate fission products and transuranics in the chloride form. Presented are preliminary results of the process qualification and characterization studies, which include chemical and physical measurements and product durability testing, of the ceramic waste form.
Date: December 5, 1997
Creator: Frank, S. M.; Bateman, K. J.; DiSanto, T.; Johnson, S. G.; Moschetti, T. L.; Noy, M. H. et al.
Partner: UNT Libraries Government Documents Department

Accelerated alpha radiation damage in a ceramic waste form, interim results

Description: Interim results are presented on the alpha-decay damage study of a {sup 238}Pu-loaded ceramic waste form (CWF). The waste form was developed to immobilize fission products and transuranic species accumulated from the electrometallurgical treatment of spent nuclear fuel. To evaluate the effects of {alpha}-decay damage on the waste form the {sup 238}Pu-loaded material was analyzed by (1) x-ray diffraction (XRD), (2) microstructure characterization by scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with energy and wavelength dispersive spectroscopy (EDS/WDS) and electron diffraction, (3) bulk density measurements and (4) waste form durability, performed by the product consistency test (PCT). While the predominate phase of plutonium in the CWF, PuO{sub 2}, shows the expected unit cell expansion due to {alpha}-decay damage, currently no significant change has occurred to the macro- or microstructure of the material. The major phase of the waste form is sodalite and contains very little Pu, although the exact amount is unknown. Interestingly, measurement of the sodalite phase unit cell is also showing very slight expansion; again, presumably from {alpha}-decay damage.
Date: November 11, 1999
Creator: Frank, S. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T. P.; Sinkler, W.; Esh, D. et al.
Partner: UNT Libraries Government Documents Department

Plutonium host phases derived from high-level waste at the Idaho Chemical Processing Plant

Description: The National Academy of Sciences recommended dissolution in a silicate matrix, with fission products to provide a protective radiation field, as one option for dispositioning excess weapons-grade plutonium. Candidate materials and processing conditions have been developed to pursue this option using high-level waste stored at the Idaho Chemical Processing Plant. Devitrification of glassy host materials achieves increased plutonium loading by partitioning plutonium into durable crystalline host phases. Results of devitrification experiments are summarized, and several unique plutonium host phases are reported. These phases were initially synthesized and characterized using cerium as a plutonium surrogate, to simplify operational requirements. Tests using plutonium were performed to validate results obtained with surrogate materials. Characterization and leach test results are reported.
Date: May 4, 1997
Creator: O`Holleran, T. P.; Johnson, S. G.; Kong, P. C. & Staples, B. A.
Partner: UNT Libraries Government Documents Department

TEM investigation of a ceramic waste form for immobilization of process salts generated during electrometallurgical treatment of spent nuclear fuel.

Description: Transmission electron microscopy (TEM) examination is presented of the microstructure of a ceramic waste form developed at Argonne National Lab - West for immobilization of actinides and fission products present in an electrorefiner salt. The material is produced by occluding the salt in zeolite granules, followed by hot isostatic pressing of the occluded zeolite in a mixture with a borosilicate glass. The paper presents results from a cold surrogate ceramic waste form, as well as {sup 239}Pu and {sup 238}Pu loaded samples.
Date: May 6, 1999
Creator: Esh, D. W.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Moschetti, T. L.; O'Holleran, T. P. et al.
Partner: UNT Libraries Government Documents Department

Study of alpha-decay damage in a glass-bonded, sodalite ceramic waste.

Description: A glass-bonded, sodalite ceramic waste form that contains fission products, uranium, and plutonium is intended for disposition in a geologic repository. Over the many years the waste is expected to be in the repository, there is a potential for waste form degradation due to alpha decay damage. To investigate the effects of alpha-decay damage in glass-bonded, sodalite ceramic waste forms, several waste forms were produced with a {sup 238}Pu loading of 1.8 weight percent. This loading is roughly ten times greater than the plutonium loading for all isotopes in the waste form intended for the repository. Due to the higher specific activity of {sup 238}Pu as well as a higher fraction of total plutonium, the same number of alpha decays per gram of material has been achieved after four years as a waste form of nominal composition after ten thousand years. This paper describes the results of different tests near the completion of a four-year study. Trends of these {sup 238}Pu-doped waste forms include volume expansion of crystalline phases and possible increases in the release rates of several elements in the chemical durability tests. There have not yet been any indications of macroscopic swelling by density measurements, amorphization by x-ray diffraction, or microstructural changes by electron microscopy. Overall, the observed changes to the waste form due to alpha-decay are not of sufficient magnitude yet to cause concern over waste form degradation.
Date: August 20, 2002
Creator: Barber, T. L.; DiSanto, T.; Frank, S. M.; Goff, K. M.; Johnson, S. G.; Jue, J.-F. et al.
Partner: UNT Libraries Government Documents Department

Plutonium-238 alpha-decay damage study of the ceramic waste form.

Description: An accelerated alpha-decay damage study of a glass-bonded sodalite ceramic waste form has recently been completed. The purpose of this study was to investigate the physical and chemical durability of the waste form after significant exposure to alpha decay. This accelerated alpha-decay study was performed by doping the ceramic waste form with {sup 238}Pu which has a much greater specific activity than {sup 239}Pu that is normally present in the waste form. The alpha-decay dose at the end of the four year study was approximately 1 x 10{sup 18} alpha-decays/gram of material. An equivalent time period for a similar dose of {sup 239}Pu would require approximately 1100 years. After four years of exposure to {sup 238}Pu alpha decay, the investigation observed little change to the physical or chemical durability of the ceramic waste form (CWF). Specifically, the {sup 238}Pu-loaded CWF maintained it's physical integrity, namely that the density remained constant and no cracking or phase de-bonding was observed. The materials chemical durability and phase stability also did not change significantly over the duration of the study. The only significant measured change was an increase of the unit-cell lattice parameters of the plutonium oxide and sodalite phases of the material and an increase in the release of salt components and plutonium of the waste form during leaching tests, but, as mentioned, these did not lead to any overall loss of waste form durability. The principal findings from this study are: (1) {sup 238}Pu-loaded CWF is similar in microstructure and phase composition to referenced waste form. (2) Pu was observed primarily as oxide comprised of aggregates of nano crystals with aggregates ranging in size from submicron to twenty microns in diameter. (3) Pu phases were primarily found in the intergranular glassy regions. (4) PuO phase shows expected unit cell volume expansion due ...
Date: March 27, 2006
Creator: Frank, S. M.; Barber, T. L.; Cummings, D. G.; DiSanto, T.; Esh, D.W.; Giglio, J. J. et al.
Partner: UNT Libraries Government Documents Department

Waste forms for plutonium disposition

Description: The field of plutonium disposition is varied and of much importance, since the Department of Energy has decided on the hybrid option for disposing of the weapons materials. This consists of either placing the Pu into mixed oxide fuel for reactors or placing the material into a stable waste form such as glass. The waste form used for Pu disposition should exhibit certain qualities: (1) provide for a suitable deterrent to guard against proliferation; (2) be of minimal volume, i.e., maximize the loading; and (3) be reasonably durable under repository-like conditions. This paper will discuss several Pu waste forms that display promising characteristics.
Date: October 1, 1997
Creator: Johnson, S.G.; O`Holleran, T.P.; Frank, S.M.; Meyer, M.K.; Hanson, M.; Staples, B.A. et al.
Partner: UNT Libraries Government Documents Department

Status of ceramic waste form degradation and radionuclide release modeling.

Description: As part of the spent fuel treatment program at Argonne National Laboratory (ANL), a ceramic waste form is being developed for disposition of the salt waste stream generated during the treatment process. Ceramic waste form (CWF) degradation and radionuclide release modeling is being carried out for the purpose of estimating the impact of the CWF on the performance of the proposed repository at Yucca Mountain. The CWF is composed of approximately 75 wt% salt-loaded sodalite encapsulated in 25 wt% glass binder. Most radionuclides are present as small inclusion phases in the glass. Since the release of radionuclides can only occur as the glass and sodalite phases dissolve, the dissolution rates of the glass and sodalite phases are modeled to provide an upper bound to radionuclide release rates from the CWF. Transition-state theory for the dissolution of aluminosilicate minerals provides a mechanistic basis for the CWF degradation model, while model parameters are obtained by experimental measurements. Performance assessment calculations are carried out using the engineered barrier system model from the Total System Performance Assessment--Viability Assessment (TSPA-VA) for the proposed repository at Yucca Mountain. The analysis presented herein suggests that the CWF will perform in the repository environment in a manner that is similar to other waste forms destined for the repository.
Date: February 26, 2003
Creator: Fanning, T. H.; Ebert, W. L.; Frank, S. M.; Hash, M. C.; Morris, E. E.; Morss, L. R. et al.
Partner: UNT Libraries Government Documents Department

Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

Description: As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.
Date: March 1, 2013
Creator: Westphal, B.R.; Marsden, K.C.; McCartin, W.M.; Frank, S.M.; D.D. Keiser, Jr.; Yoo, T.S. et al.
Partner: UNT Libraries Government Documents Department

Corrosion tests with uranium- and plutonium-loaded ceramic waste forms.

Description: Tests were conducted with ceramic waste form (CWF) materials that contained small amounts of uranium and plutonium to study their release behavior as the CWF corroded. Materials made using the hot isostatic press (HIP) and pressureless consolidation (PC) methods were examined and tested. Four different materials were made using the HIP method with two salts having different U:Pu mole ratios and two zeolite reagents having different residual water contents. Tests with the four HIP U,Pu-loaded CWF materials were conducted at 90 and 120 C, at CWF-to-water mass ratios of 1:10 and 1:20, and for durations between 7 and 365 days. Materials made using two PC processing conditions were also tested. Tests with the two PC U,Pu-loaded CWF materials were conducted at 90 and 120 C, at a CWF-to-water mass ratio of 1:10, and for durations between 7 and 182 days. The releases of matrix elements, U, and Pu in tests conducted under different test conditions and with different materials are compared to evaluate the effects of composition and processing conditions on the release behavior of U and Pu and the chemical durabilities of the different materials. The distributions of released elements among the fractions that were dissolved, in colloidal form in the solution, and fixed to test vessel walls were measured and compared. Characterization of Pu-bearing colloidal particles recovered from the test solutions using solids analysis techniques are also reported. The principal findings from this study are: (1) The release of U and Pu is about 10X less than the release of Si and 50X less than the release of B under all test conditions. This implies that U and Pu are in a phase that is less soluble than the sodalite and binder glass matrix. (2) Almost all of the plutonium that is released from U,Pu-loaded CWF is present ...
Date: January 9, 2003
Creator: Morss, L. R.; Johnson, S. G.; Ebert, W. L.; DiSanto, T.; Frank, S. M.; Holly, J. L. et al.
Partner: UNT Libraries Government Documents Department