Low-level Mixed Waste Research Articles

Chemical Characteristics of Leachate in Low-Level Radioactive Waste Disposal Facilities

AbstractLeachate data from four low-level radioactive waste and mixed waste disposal facilities operated by the U.S. DOE for environmental restoration programs are presented and analyzed. Constitue...

Decontamination and Deactivation of the Power Burst Facility at the Idaho National Laboratory

Successful decontamination and deactivation of the Power Burst Facility located at the Idaho National Laboratory was accomplished through the use of extensive planning, job sequencing, engineering controls, continuous radiological support, and the use of a dedicated group of experienced workers. Activities included the removal and disposal of irradiated fuel, miscellaneous reactor components and debris stored in the canal, removal and disposition of a 15.6 curie Pu:Be start-up source, removal of an irradiated in-pile tube, and the removal of approximately 220,000 pounds of lead that was used as shielding primarily in Cubicle 13. The canal and reactor vessel were drained and water was transferred to an evaporation tank adjacent to the facility. The canal was decontaminated using underwater divers, and epoxy was affixed to the interior surfaces of the canal to contain loose contamination. The support structures and concrete or steel frame walls that form the confinement were left in place. The reactor core was left in place and a carbon steel shielding plate was placed over the reactor core to reduce radiation levels. All low-level waste and mixed low level waste generated as a result of the work activities was characterized and disposed.

Suspensions of fluor-containing nanoparticles for quantifying β−-emitting radionuclides in non-hazardous media

Nanoparticles containing scintillants were prepared from oil-in-water microemulsions in which styrene-divinylbenzene comprised the oil phase. Primary and secondary fluors [2,5-diphenyloxazole and 1,4-bis(2-methylstyryl) benzene] were dissolved in the oil phase, which was emulsified with dodecyl trimethylammonium bromide or sodium dodecyl sulfate before polymerization by the addition of sodium persulfate. The resulting suspension qualified as non-hazardous waste by testing against US EPA standards for reactivity, pH, ignitability, and toxicity characteristic leaching.14C-glycerol was added to the aqueous nanoparticle suspension and to a conventional organic scintillation cocktail, and the relative efficiency of detection was obtained by liquid scintillation counting. The results were compared with a mathematical model that predicted the effect of particle size and concentration on the probability of an emitted β− particle interacting with a suspended particle. The use of these nanoparticle suspensions represents an opportunity to reduce mixed low-level radioactive hazardous waste generated by liquid scintillation counting.

Kinetics of Mixed Low-Level Waste Incapsulation Using Iron Phosphate Chemically Bonded Cement

ABSTRACTThe iron phosphate cement (IPC) structure formation process has been studied by Mossbauer spectroscopy. The IPC is formed as a result of interaction of iron oxides with orthophosphoric acid and could be used as a matrix for immobilization of low-level radioactive waste.The structural formation process has been shown may be considered to consist of two stages. The first stage lasts to the moment of setting, when the diffusion process goes on in the liquid phase. At this stage the main fragments of the structural polymeric frame of the IPC are developed consisting of iron (+2) and (+3) phosphates. The rate of the oxides-with-phosphoric-acid reaction as well as the time of hardening depends on the bivalent iron content.The second stage begins from the moment of setting when the diffusion becomes slower. At this stage the process is characterized by the negligible increase in the iron phosphates (+2), (+3) content and transformation of the previously formed phosphates.The nuclear gamma-resonance (NGR) parameters have been determined of FeH3(PO4)2·2.5H2O, forming in the hematite (Fe2O3) based IPC: isomeric shift (IS) = 0.46 mm/s, quadrupole splitting (QS) = 0.197 mm/s, FWHM =0.282 mm/s.

High-resolution stratigraphic modeling of the vadose zone at the Savannah River Site low-level radioactive waste trenches disposal facility

The principal disposal facility for solid, hazardous, and mixed low-level radioactive wastes is located in the central part of the Savannah River Site. These wastes are generally disposed of in engineered, unlined trenches and excavations. The trenches vary in length and width but are typically constructed in the upper 7–8 m (21–24 ft) of the local sediments and in the vadose zone. Water from the surface must pass through the waste trenches and unsaturated vadose zone to reach the saturated water table and may contribute to groundwater contamination. The vadose zone underlying the study area extends to depths between approximately 20 and 24 m (65 and 78 ft). To comply with regulatory requirements, an investigation program was developed to characterize the vadose zone underlying two active disposal trench areas. A high-resolution vadose zone lithostratigraphic model was required before the installation and placement of experimental equipment to measure soil hydraulic properties. A detailed mathematical model of flow and transport could then use the stratigraphic model and the water flux information from the instrumentation, for example, to constrain a finite-element grid and arrive at the contaminant flow characteristics in the vadose interval. The two high-resolution stratigraphic models were developed and interpreted from field and laboratory data collected from continuous split-spoon borings, Shelby tube soil samples, piezocone penetration soundings using cone penetrometer technology (CPT), and borehole geophysical logs from nearby shallow monitoring wells. Exposed stratigraphy in open trenches and nearby cuts was also used. These models revealed (1) detailed lithostratigraphic and soil moisture conditions in the vadose zone, including three variations in depositional character roughly corresponding to previously defined formational contacts; (2) that indications that the lateral extent of many vadose zone bedforms is on the order of 20 m (66 ft) or less, therefore requiring vadose zone sampling points to be placed on spacings of less than 20 m (66 ft) for modeling; and (3) that once correlation and modeling have been performed, it is possible to use CPT data, matched to the model, to precisely place new sensors and instrument.

Managing a low-level mixed waste storage facility: a checklist for compliance to 40 CFR 266.

Do you hold radioactive material for storage and decay before disposal? Does your material contain any hazardous chemical as defined by the EPA? You may need to take a closer look at the Final Rule published by the EPA in the May 16, 2001 Federal Register.

Performance Analysis for Disposal of Mixed Low-Level Waste. I: Methodology

A simple methodology has been developed for evaluating the technical capabilities of potential sites for disposal of mixed low-level radioactive waste. The results of the evaluation are expressed as “permissible radionuclide concentrations in disposed waste.” The methodology includes an analysis of three separate pathways: (1) releases of radionuclides to groundwater; (2) releases of potentially volatile radionuclides to the atmosphere; and (3) the consequences of inadvertent intrusion into a disposal facility. For each radionuclide, its limiting permissible concentration in disposed waste is the lowest of the permissible concentrations determined from each of the three pathways. These permissible concentrations in waste at an evaluated site can be used to assess the capability of the site to dispose of waste streams containing multiple radionuclides.

Performance Analysis for Disposal of Mixed Low-Level Waste. II: Results

Ten sites were evaluated as potential locations for disposal of mixed low-level radioactive waste using a simple methodology. Three environmental pathways (water, atmospheric, and inadvertent intruder) were analyzed using generic transport models that were modified for site-specific conditions. The results were summarized by grouping the 58 evaluated radionuclides according to their half-lives and environmental mobility and by their limiting pathway (i.e., the pathway providing the lowest permissible radionuclide concentration in disposed waste of the three evaluated pathways). The results indicate that all evaluated sites have the technical capability for disposal of some radionuclides in the waste. For most radionuclides, the intruder scenarios were more important in determining permissible radionuclide concentrations than the other pathways, particularly for arid sites. For humid sites, if the water pathway is not the most limiting and the permissible radionuclide concentration is high, a more sophisticated and rigorous analysis of the water pathway may not be warranted. However, if the permissible concentration is relatively low, more refined analyses may produce higher permissible concentrations based on additional site characterization data.

Partitioning of plutonium between glass, secondary phases, and contact refractories

AbstractA wide variety of plutonium-containing wastes have been generated from both defense and commercial nuclear programs. The Advanced Mixed Waste Treatment Project (AMWTP) involves the privatized treatment of 65,000 m3 of mixed TRU and low-level waste that is in storage at DOE's INEEL site. BNFL Inc. leads the commercial team including GTS Duratek, Inc., that has been selected to design, construct, and operate a facility to perform that service. Since vitrification is a potential treatment option, the current study was conducted to investigate issues surrounding the incorporation of the approximately 700 kg of plutonium that is present in the AMWTP waste. A series of glasses were prepared with plutonium (0.5 wt%), with cerium (as a potential surrogate), with both, and with neither. Scanning electron microscopy coupled with energy dispersive x-ray spectroscopy (SEM-EDS) and optical analyses confirmed that both elements were completely dissolved into the as-melted glasses. Heat treatments forced the formation of secondary phases, which were analyzed for Pu and Ce incorporation. Plutonium-containing glasses were spiked with sulfates and remelted to force the formation of a sulfate salt phase, which was analyzed to determine partitioning of Pu and Ce. Coupons of Monofrax K-3 glass contact refractory were exposed to melts of these glasses for six days at 1208°C using both standard and modified versions of ASTM procedure C621-84. Thickness loss, microstructural alteration, and the extent to which Pu and Ce were incorporated into the material were then determined by methods employing SEMEDS\ wavelength dispersive x-ray spectroscopy (WDS). Extreme heat treatments were found to lead to the formation of small amounts (< 0.1 vol%) of either (Pu, Ce)O2, or PuO2, depending upon whether Ce was present or absent, respectively. The Pu level in the sulfate phase was at least a factor often below that of the parent glass, and no tendency to concentrate Pu in the K-3 refractory above the level in the parent glass was found.

Vitrified Waste Form Performance Modeling Applied to the Treatment and Disposal of a Mixed-Waste Sludge at the Savannah River Site

Vitrification, the conversion of source components into a solid amorphous glass matrix, has emerged as a viable treatment technology for low-level radioactive waste and mixed waste. To dispose of vitrified low-level waste at U.S. Department of Energy facilities, site-specific radiological performance assessments must be conducted to demonstrate that waste glass satisfies performance objectives for environmental protection. More than 2500 m3 of F006-listed low-level mixed-waste sludge stored in the Reactor Materials Department (M-Area) at the Savannah River Site (SRS) is scheduled for vitrification. This study evaluates the feasibility of on-site disposal of vitrified M-Area waste at SRS.Laboratory leaching tests that accelerate the glass corrosion process are currently the best indicators of vitrified waste form durability. A method to incorporate laboratory leaching data into performance assessments is presented. A screening-level performance assessment code is used to model trench disposal of M-Area waste glass. The allowable leach rate for vitrified M-Area waste is determined based on both a maximum radiological dose equivalent of 4 mrem/yr for the drinking water pathway and a maximum uranium concentration of 20 μg/l in groundwater. The allowable leach rate is compared with published long-term leaching data for a wide range of waste glass compositions and test conditions. This analysis demonstrates that trench disposal of the waste glass is likely to meet applicable performance objectives if the glass is of above average durability compared with the reference set of glasses.

Management and disposal of waste from sites contaminated by radioactivity

Various methods of managing and disposing of wastes generated by decontamination and decommissioning (D & D) activities are described. This review of current waste management practices includes a description of waste minimization and volume reduction techniques and their applicability to various categories of radwaste. The importance of the physical properties of the radiation and radioactivity in determining the methodology of choice throughout the D & D process is stressed. The subject is introduced by a survey of the common types of radioactive contamination that must be managed and the more important hazards associated with each type. Comparisons are made among high level, transuranic, low level, and radioactive mixed waste, and technologically-enhanced, naturally-occurring radioactive material (TENORM). The development of appropriate clean-up criteria for each category of contaminated waste is described with the aid of examples drawn from actual practice. This includes a discussion of the application of pathway analysis to the derivation of residual radioactive material guidelines. The choice between interim storage and permanent disposal of radioactive wastes is addressed. Approaches to permanent disposal of each category of radioactive waste are described and illustrated with examples of facilities that have been constructed or are planned for implementation in the near future. Actual experience at older, existing, low-level waste disposal facilities is discussed briefly.

Phosphate ceramic process for macroencapsulation and stabilization of low-level debris wastes

Across the DOE complex, large quantities of contaminated debris and irradiated lead bricks have been accumulated for disposal. Under the US Environmental Protection Agency’s Alternative Treatment Standards, the preferred method of disposal of these wastes is macroencapsulation. Chemically bonded phosphate ceramic (CBPC) is a novel binder that was developed at Argonne National Laboratory to stabilize and solidify various low-level mixed wastes. This binder is extremely strong, dense, and impervious to water. In this investigation, CBPC has been used to demonstrate macroencapsulation of various contaminated debris wastes, including cryofractured debris, lead bricks, lead-lined plastic gloves, and mercury-contaminated crushed glass. This paper describes the fabrication of the waste forms, as well as the results of various characterizations performed on the waste forms. The results show that the simple and low-cost CBPC is an excellent material system for the macroencapsulation of debris wastes.

Volumetric and radiological capacity requirements for disposal of treated mixed low-level waste

The U.S. Department of Energy (DOE) has stored or expects to generate over the next five years more than 130,000 m 3 of mixed low-level waste (MLLW). Before disposal, MLLW is usually treated to comply with the land disposal restrictions of the Resource Conservation and Recovery Act. Depending on treatment type, the volumes and radionuclide concentrations of the waste streams may change. These changes must be taken into account in determining necessary disposal capacity at a site. Treatment may remove the characteristic feature that caused classification as mixed waste. Treatment may also, by reduction of mass, increase the concentrations of transuranic radionuclides sufficiently so that some waste becomes transuranic waste. This paper describes an analysis of the DOE MLLW streams to determine treated volumes and radionuclide concentrations. The waste streams were reclassified as residual MLLW, or low-level or transuranic waste resulting from treatment. About 89,000 m 3 of waste was identified for disposal as residual MLLW. Fifteen DOE sites were evaluated to determine their capabilities for hosting disposal facilities for residual MLLW. Waste streams associated with about 90% of the residual MLLW volume are likely to present no significant issues for disposal. Future studies should focus on the remaining waste streams that are potentially problematic by examining site-specific waste acceptance criteria, alternative treatment processes, alternative waste forms for disposal, and pending changes in regulatory requirements.

Characterization of low-level mixed waste sludges: Identifying and controlling sources of variability

Over the course of several years, we have examined the viability of treating low-level mixed wastes from several US Department of Energy (DOE) sites by vitrification. In particular, extensive studies have been performed on actual low-level mixed wastes from the Fernald site in Ohio. We have shown that it is possible to produce durable glasses (pass the US EPA TCLP test and perform comparably to high-level waste glasses on the PCT test) with waste loadings of 80–98%. The results of these vitrification studies have been used to up-scale from laboratory scale melters to a demonstration-scale melter operating at the Fernald site with a nominal glass production rate of 300 kg/day. This paper focuses on the difficulties that we have encountered in characterizing and working with actual waste sludges and the solutions that we have developed.

Characteristics of Materials Produced by Plasma Treatment of Radioactive Wastes

ABSTRACTPlasma processes become more and more attractive for the treatment of radioactive low/medium level waste. Plants with throughput rates ranging from 20 kilograms to several hundred kilograms per hour are under construction or under evaluation in different countries. One major installation is under construction at the ZWILAG-site in Wurenlingen (Switzerland). ZWILAG (Zwischenlager Wurenlingen AG) is a company being responsible for treatment and intermediate storage of all radioactive wastes occurring in Switzerland. Thermal destruction (pyrolysis) of organic material and melting of the inorganic residuals and inorganic feed stock can be done simultaneously in a plasma heated vessel. The following output materials are obtained: Metal, glassy slag, off-gas. Depending on feed stock and storage requirements these outputs can have different properties. The distribution of the activity is of particular importance. Isotopes like cobalt-60 (as a metal) partition almost quantitatively to the metal phase. Volatile isotopes like cesium-137 partition to some extent to the off-gas, they can however be retained in the slag to a high extent by employing a two stage process. For some applications the activity in the metal can be kept below the activity tolerance limits making the metal suitable for even non nuclear applications. The most important property of the slag is its leaching behaviour. Mechanical properties can be of importance, too. The paper will present examples from different types of radioactive waste: typical low level mixed waste, ion exchange resins and zeolites, mixtures of concrete and metal, contaminated soil, material contaminated with natural activity and others. Composition of the materials, activity distributions, data on leaching behaviour and some mechanical properties are presented

Remediation of low-level mixed waste: cellulose-based materials and plutonium

Low-level mixed radioactive wastes containing cellulose-based materials and plutonium have been generated during various nuclear processing activities. Biological digestion of the organic- or cellulose- based material was examined as an environmentally acceptable and effective method of treatment for these and other similar wastes. Cellulase enzyme was used to initiate biodegradation prior to 90% destruction of the cellulose material by a sewage sludge consortium. Plutonium did not significantly effect the biodegradation. Bench-scale experimental data were used to design a batch treatment system. A cost and sensitivity analysis was performed to determine the optimal reactor size, materials of construction and media type. The sensitivity analysis indicated that while a 12-month treatment scenario using a carbon steel ball mill, sludge digester and vacuum thickener was the least expensive scenario evaluated on a levelized cost basis ($800 per ton of waste degraded per month), the 12-month scenario using stainless steel construction and the alternative dewatering system offered the most cost-effective treatment alternative and better corrosion resistance (levelized cost of $1130 per ton per month). The dewatering system consisting of a disk centrifuge and sludge dryer is capable of doubling the sludge solids content and produce an overall waste reduction of 67%. The proposed waste treatment system offers a cost savings of up to 31% compared to conventional disposal practices.

Low temperature incineration of mixed wastes using bulk metal oxide catalysts

Volume reduction of low-level mixed wastes from former nuclear weapons facilities is a significant environmental problem. Processing of these materials presents unique scientific and engineering problems due to the presence of minute quantities of radionuclides which must be contained and concentrated for later safe disposal. Low-temperature catalytic incineration is one option that has been utilized at the Rocky Flats facility for this purpose. This paper presents results of research regarding evaluation of bulk metal oxides as catalysts for low-temperature incineration of carbonaceous residues which are typical by-products of fluidized bed combustion of mixed wastes under oxygen-lean conditions. A series of 14 metal oxides were screened in a thermogravimetric analyzer, using on-line mass spectrometry for speciation of reaction product gases. Catalyst evaluation criteria focused on the thermal—redox activity of the metals using both carbon black and PVC char as surrogate waste materials. Results indicated that metal oxides which were P-type semiconductor materials were suitable as catalysts for this application. Oxides of cobalt, molybdenum, vanadium, and manganese were found to be particularly stable and active catalysts under conditions specific to this process ( T < 650°C, low oxygen partial pressures). Bench-scale evaluation of these metal oxides with respect to stability to chlorine (HCl) attack was carried out at 550°C using a TG/MS system. Cobalt oxide was found to be resistant to metal loss in a HCl/He gaseous environment while metal loss from Mo, Mn, and V-based catalysts was moderate to severe. XRD and SEM/EDX analysis of spent Co catalysts indicated the formation of non-stoichiometric cobalt chlorides. Regeneration of chlorinated cobalt was found to successfully restore the low-temperature combustion activity to that of the fresh metal oxide.

Benzene destruction in aqueous waste—I. Bench-scale gamma irradiation experiments

Destruction of the benzene component of a simulated low-level mixed aqueous waste stream by high energy irradiation was explored. This work was motivated by the fact that mixed waste, containing both radionuclides and regulated (non-radioactive) chemicals, is more difficult and more expensive to dispose of than only radioactive waste. After the benzene is destroyed, the waste can then be listed only as radiological waste instead of mixed waste, simplifying its disposal. This study quantifies the removal of benzene, and the formation and destruction of reaction products in a relatively complex waste stream matrix consisting of NO 3 −, SO 4 2−, PO 4 3−, Fe 2+ and detergent at a pH of 3. All of the experiments were conducted at a bench scale using a 60Co gamma source.

Tritium Waste Management for TFTR D-T Operations

This paper describes the low level radioactive and mixed wastes generated at the Princeton Plasma Physics Laboratory (PPPL) by the Tokamak Fusion Test Reactor (TFTR) Deuterium-Tritium (D-T) operations. The D-T operations are in progress and will continue through most of 1995. The data presented is for the wastes generated at the TFTR facility in calendar year (CY) 1994.

95/05420 Total caesium-fixing potentials of acid organic soils

Various methods of managing and disposing of wastes generated by decontamination and decommissioning (D & D) activities are described. This review of current waste management practices includes a description of waste minimization and volume reduction techniques and their applicability to various categories of radwaste. The importance of the physical properties of the radiation and radioactivity in determining the methodology of choice throughout the D & D process is stressed.The subject is introduced by a survey of the common types of radioactive contamination that must be managed and the more important hazards associated with each type. Comparisons are made among high level, transuranic, low level, and radioactive mixed waste, and technologically-enhanced, naturally-occurring radioactive material (TENORM). The development of appropriate clean-up criteria for each category of contaminated waste is described with the aid of examples drawn from actual practice. This includes a discussion of the application of pathway analysis to the derivation of residual radioactive material guidelines.The choice between interim storage and permanent disposal of radioactive wastes is addressed. Approaches to permanent disposal of each category of radioactive waste are described and illustrated with examples of facilities that have been constructed or are planned for implementation in the near future. Actual experience at older, existing, low-level waste disposal facilities is discussed briefly.