Low-activity Waste Research Articles

Pre-viking Swedish hillfort glass: A prospective long-term alteration analogue for vitrified nuclear waste.

Models for long-term glass alteration are required to satisfy performance predictions of vitrified nuclear waste in various disposal scenarios. Durability parameters are usually extracted from short-term laboratory tests, and sometimes checked with long-term natural experiments on glasses, termed analogues. In this paper, a unique potential ancient glass analogue from Sweden is discussed. The hillfort glass found at Broborg represents a unique case study as a vitrified waste glass analogue to compare to Low Activity Waste glass to be emplaced in near surface conditions at Hanford (USA). Glasses at Broborg have similar and dissimilar compositions to LAW glasses, allowing the testing of long-term alteration of different glass chemistries. In addition, the environmental history of the site is reasonably well documented. Initial investigations on previously collected samples established methodologies for handling and characterizing these artifacts by laboratory methods while preserving their alteration layers and cultural context. Evidence of possible biologically influenced glass alteration, and differential alteration in the 2 types of glass found at the Broborg site is presented.

Effect of Technetium-99 sources on its retention in low activity waste glass

Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO2∙2H2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.

Melter feed viscosity during conversion to glass: Comparison between low‐activity waste and high‐level waste feeds

AbstractDuring nuclear waste vitrification, a melter feed (a slurry mixture of a nuclear waste and various glass forming and modifying additives) is charged into the melter where undissolved refractory constituents are suspended together with evolved gas bubbles from complex reactions. Knowledge of flow properties of various reacting melter feeds is necessary to understand their unique feed‐to‐glass conversion processes occurring within a floating layer of melter feed called a cold cap. The viscosity of two low‐activity waste (LAW) melter feeds were studied during heating and correlated with volume fractions of undissolved solid phase and gas phase. In contrast to the high‐level waste (HLW) melter feed, the effects of undissolved solid and gas phases play comparable roles and are required to represent the viscosity of LAW melter feeds. This study can help bring physical insights to feed viscosity of reacting melter feeds with different compositions and foaming behavior in nuclear waste vitrification.

Tellurite glasses for vitrification of technetium-99 from pyrochemical processing

A new alkali-alumino tellurite glass composition was developed to immobilize highly-volatile technetium (Tc) wastes generated from the pyrochemical processing technology. Tellurite glass can incorporate up to 7 mass% of rhenium (Re, used as a surrogate for Tc) with an average retention of 86%. Normalized elemental releases evaluated by seven-day product consistency test (PCT) satisfied the immobilized low activity waste requirements of United States when concentration of Ca(ReO4)2 in the glass was <12 mass%. Re ions form Re7+ and are coordinated with four oxygens to form ReO4− tetrahedra. These tetrahedra bond to modifiers such as Ca2+ or Na+ that are further connected to the tellurite glass network by Ca2+ (or Na+) – non-bridging oxygen bonds.

Tc removal from the waste treatment and immobilization plant low-activity waste vitrification off-gas recycle

ABSTRACTVitrification of Low Activity Waste in the Hanford Waste Treatment and Immobilization Plant generates a condensate stream from the off-gas processes. Components in this stream are partially volatile and accumulate to high concentrations through recycling, which impacts the waste glass loading and facility throughput. The primary radionuclide that vaporizes and accumulates in the stream is 99Tc. This program is investigating Tc removal via reductive precipitation with stannous chloride to examine the potential for diverting this stream to an alternate disposition path. Research has shown stannous chloride to be effective, and this paper describes results of recent experiments performed to further mature the technology.

Removal of radiocesium from low level radioactive effluents by hexacyanoferrate loaded synthetic zeolite: laboratory to pilot plant scale demonstration

Abstract Present paper reports removal of radiocesium from low level waste using a modified sorbent (13X-CFC) prepared by in-situ precipitation of potassium copper hexacyanoferrate(II) inside the macropores of a synthetic zeolite. The Cs exchange isotherm of the sorbent is established and it found to follow Fruendlich absorption isotherm equation. It is varified that presence of hexacyanoferrate on zeolite facilitates rapid Cs uptake performance. This is further confirmed in laboratory scale column tests, wherein excellent Cs removal performance from low level waste simulant was observed even at higher flow rates (40 bed volumes per hour). The utility of the sorbent is established through successful demonstration in a pilot scale (50 L) trial with almost complete removal of 137Cs from more than 14,000 bed volumes of actual low level waste. The sorbent, owing to its low cost and excellent 137Cs removal performance, is expected to find application in treatment of very low active waste streams.

Iron phosphate glass waste forms for vitrifying Hanford AZ102 low activity waste (LAW), part I: Glass formation model

A methodology for determining glass formation in a 5-component iron phosphate base glass system that contained P2O5, Fe2O3, Al2O3, Na2O and SO3 has been developed using a standard 5-component D-Optimal Statistical Mixture Design procedure. This compositional system was selected in order to study the feasibility of vitrifying a high soda (~80wt%), a high sulfate (~17wt%) containing Hanford AZ102 LAW in iron phosphate glasses with waste loadings much higher than the currently reported values. The statistical analysis was designed to develop a series of initial compositions (total 28), nearly half of which formed glass upon melting and cooling. After separating the compositions that formed glass from those that crystallized, three molar ratios, namely, (O/Na), (Na/Al) and (Na/P), were identified as potential parameters for defining glass formation (or crystallization). The (O/Na) molar ratio had the greatest effect. Melts whose (O/Na), (Na/P) or (Na/Al) molar ratios were ≤4.35, ≥1.13, or ≥2.56, respectively, had the highest likelihood to form glass when cooled. The melts crystallized completely when the values of these three molar ratios were, respectively, ≥6.02, ≤0.77 or ≤1.92. For the compositions that formed glass, the waste loadings as defined by the sum of Na2O and SO3, the two major components comprising ˃90wt% of the Hanford AZ102 LAW, ranged from 25 to 33wt%.A glass formation-composition model that formalized the above findings was developed using the Classification and Regression Tree (CART) analyses. The CART model was validated by testing the glass formation characteristics of several additional compositions whose (O/Na) molar ratio was maintained ≤4.35, i.e., the compositions were expected to form glass. This model is expected to identify similar molar ratios for predicting glass formation/crystallization of iron phosphate based compositions containing not only other Hanford LAW streams, but high level wastes as well. Using the compositions that formed glass, property-composition statistical models for the waste forms were also developed and will be reported in a second paper.

The function of Sn(II)-apatite as a Tc immobilizing agent

At the U.S. Department of Energy Hanford Site, Tc-99 is a component of low-activity waste (LAW) fractions of the nuclear tank waste and removal of Tc from LAW streams would greatly benefit the site remediation process. In this study, we investigated the removal of Tc(VII), as pertechnetate, from deionized water (DIW) and a LAW simulant through batch sorption testing and solid phase characterization using tin (II) apatite (Sn-A) and SnCl2. Sn-A showed higher levels of Tc removal from both DIW and LAW simulant. Scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/XEDS) and X-ray absorption spectroscopy (XAS) of reacted Sn-A in DIW showed that TcO4- is reduced to Tc(IV) on the Sn-A surface. The performance of Sn-A in the LAW simulant was lowered due to a combined effect of the high alkalinity, which lead to an increased dissolution of Sn from the Sn-A, and a preference for the reduction of Cr(VI).

Silver-based getters for 129I removal from low-activity waste

Abstract A prominent radionuclide of concern in nuclear wastes, 129I, is present in low-activity wastes (LAW) at the Hanford site. Several Ag-containing materials were tested as immobilization agents, or “getters”, for I (as iodide, I−) removal from deionized (DI) water and a liquid LAW simulant: Ag impregnated activate carbon (Ag–C), Ag exchanged zeolite (Ag–Z), and argentite. In anoxic batch experiments with DI water, the Ag–C and argentite were most effective, with maximum Kd values of 6.2 × 105 mL/g for the Ag–C and 3.7 × 105 mL/g for the argentite after 15 days. Surface area and Ag content were found to influence the performance of the getters in DI water. In the anoxic batch experiments with LAW simulant, Ag–Z vastly outperformed the other getters with Kd values of 2.2 × 104 mL/g at 2 h, which held steady until 15 days, compared with 1.8 × 103 mL/g reached at 15 days by the argentite. All getters were stable over long periods of time (i.e. 40 days) in DI water, while the Ag–Z and argentite were also stable in the LAW simulant. Ag–Z was found to have consistent I removal upon crushing to a smaller particle size and in the presence of O2, making it a strong candidate for the treatment of LAW containing I.

Formation of Technetium Salts in Hanford Low‐Activity Waste Glass

The distribution and physical form of technetium in a Hanford low‐activity waste (LAW) glass was examined with scanning electron microscopy (SEM) and X‐ray diffraction (XRD). A simulated Hanford LAW glass was spiked with varying amounts of potassium pertechnetate and melted at 1000°C. The glass was melted in a sealed quartz ampoule with the air pumped out, so that volatile material could leave the glass but would not be lost from the system. Previous studies have shown that technetium remains in the glass up to about 2000 ppm, but rises to the top of the melt as a separate salt phase above this concentration. Examination by SEM shows that crystals of technetium compounds appear to grow out of the hot glass, which implies that the hot glass was supersaturated in technetium salts. Some of the technetium compound crystals had apparently melted, but other crystals had obviously not melted and must have formed after the glass had partially cooled. The technetium compounds in the salt layer are KTcO4 and NaTcO4, according to SEM and XRD. No TcO2 was found in the salt phase, even though Tc(IV) has been previously reported in the glass.

Sodalite as a vehicle to increase Re retention in waste glass simulant during vitrification

Technetium (Tc) retention during Hanford waste vitrification can be increased if the volatility can be controlled. Incorporating Tc into a thermally stable mineral phase, such as sodalite, is one way to achieve increased retention. Here, rhenium (Re)-bearing sodalite was tested as a vehicle to transport perrhenate (ReO4−), a nonradioactive surrogate for pertechnetate (TcO4−), into high-level (HLW) and low-activity waste (LAW) glass simulants. After melting HLW and LAW simulant feeds, the retention of Re in the glass was measured and compared with the Re retention in glass prepared from a feed containing Re2O7. Phase analysis of sodalite in both these glasses across a profile of temperatures describes the durability of Re-sodalite during the feed-to-glass transition. The use of Re sodalite improved the Re retention by 21% for HLW glass and 85% for LAW glass, demonstrating the potential improvement in Tc-retention if TcO4− were to be encapsulated in a Tc-sodalite prior to vitrification.

Removal of TcO4– from Representative Nuclear Waste Streams with Layered Potassium Metal Sulfide Materials

Many efforts have focused on the sequestration and immobilization of 99Tc because the radionuclide is highly mobile in oxidizing environments and presents serious health risks due to its radiotoxicity and long half-life (t1/2 = 213 000 a). One of the more common methods for Tc removal from solution and immobilization in solids is based on reducing Tc from highly soluble Tc(VII) to sparingly soluble Tc(IV). Here, we report results obtained with two potassium metal sulfides (KMS-2 and KMS-2-SS) that are capable of reducing Tc(VII) to Tc(IV). Batch sorption experiments were performed in both oxic and anoxic conditions for 15 d in both deionized water (DIW) and a highly caustic (pH ∼ 13.6), high ionic strength (8.0 mol L–1), low-activity waste (LAW) stream simulant solution. Tc removal for both materials in DIW is improved in anoxic conditions compared to oxic conditions as a result of a higher solution pH. In DIW and anoxic conditions, KMS-2 is capable of removing ∼45% of Tc, and KMS-2-SS is capable of remov...

Development of a purification technology for treatment of medium- and low-activity radioactive waste of radiochemical production from Co-60 and Cs-137

The technological flowchart of purification of medium- and low-activity waste from Co-60 and Cs-137 is developed and introduced. The developed purification scheme has been successfully tested using genuine medium- and low-level liquid radioactive waste of radiochemical production containing complexing and colloid forming components complexons, surfactants. The optimal conditions of the presented method of purification ensure reduction of the residual specific activity of 60Co and 137Cs radionuclides in the solution to less than 0,9 Bq per litre.

Characterization of SuperLig® 639 Rhenium and Technetium Resin with Batch Contact and Column Tests

Two new (2013) lots of SuperLig® 639 ion exchange resin (IBC Advanced Technologies, American Forks, UT) were tested for the first time above typical sodium concentrations (7.8 M sodium, along with typical 5 M concentration) in highly alkaline solutions. Batch contact and ion exchange column tests characterized rhenium (perrhenate) adsorption as a surrogate for pertechnetate. The work supports technetium removal options for Supplemental Low Activity Waste processing at the Hanford River Protection Project Waste Treatment Plant (WTP). The current work found that the resin performs well in the 7.8 M sodium simulant despite complete floating of the beads. A notable difference in performance between the two new resin lots was found. Resin loading overall versus temperature, potassium concentration, and rhenium/nitrate ratios is consistent with previous data and expectations despite the high sodium concentration and floating of the resin beads.

Reactions during melting of low-activity waste glasses and their effects on the retention of rhenium as a surrogate for technetium-99

Volatile loss of radioactive technetium-99 (99Tc) to off-gas is a major challenge when vitrifying low-activity waste (LAW) at the U.S. Department of Energy's Hanford Site in Washington State. We investigated the partitioning and incorporation of rhenium (Re) (a nonradioactive surrogate for 99Tc) into the glass melt during crucible melting of two simulated LAW feeds that have exhibited a large difference in 99mTc/Re retention in glass from small-scale melter tests. Each feed was prepared from a simulated liquid LAW and additives (boric acid, silica sand, etc.). The as-mixed slurry feeds were dried at 105°C and heated to 600–1100°C at 5K/min. The dried feeds and heat-treated samples were leached with deionized water for 10min at room temperature followed by 24-h leaching at 80°C. Chemical compositions of the resulting solutions and insoluble solids were analyzed. Volume expansion measurements and X-ray diffraction (XRD) analyses were performed on dried feeds and heat-treated samples to characterize the progress of feed-to-glass conversion reactions. We found that incorporation of Re into the glass melt was virtually completed during the major feed-to-glass conversion reactions that occurred at ≤700°C. The results of our study suggest that the different compositions of the salt phases formed during early stages of melting at ≤700°C are responsible for the large difference in Re incorporation into the glass melt in these two feeds.

The nature of the volatile technetium species formed during vitrification of borosilicate glass

Vitrification of sodium pertechnetate into borosilicate glass was performed in air at 1100 °C. A glass with a composition similar to the one developed for vitrification of the low activity waste at the Hanford site was used. A red volatile species was observed above 600 °C. The extended X-ray absorption fine structure results indicate the environment of the absorbing Tc atom consists of 2.9(6) O atoms at 1.73(2) A, 2.2(4) O atoms at 2.02(2) A, and 0.8(2) O atoms at 2.18(2) A. The results are consistent with the presence of a mononuclear species with a structure closely related to TcO3(OH)(H2O)2.

Rhenium volatilization in waste glasses

We investigated volatilization of rhenium (Re), sulfur, cesium, and iodine during the course of conversion of high-level waste melter feed to glass and compared the results for Re volatilization with those in low-activity waste borosilicate glasses. Whereas Re did not volatilize from high-level waste feed heated at 5Kmin−1 until 1000°C, it began to volatilize from low-activity waste borosilicate glass feeds at ∼600°C, a temperature ∼200°C below the onset temperature of evaporation from pure KReO4. Below 800°C, perrhenate evaporation in low-activity waste melter feeds was enhanced by vigorous foaming and generation of gases from molten salts as they reacted with the glass-forming constituents. At high temperatures, when the glass-forming phase was consolidated, perrhenates were transported to the top surface of glass melt in bubbles, typically together with sulfates and halides. Based on the results of this study (to be considered preliminary at this stage), the high-level waste glass with less foaming and salts appears a promising medium for technetium immobilization.

Glass Property Models, Constraints, and Formulation Approaches for Vitrification of High-Level Nuclear Wastes at the US Hanford Site

Current plans for legacy nuclear wastes stored in underground tanks at the U.S. Department of Energy’s Hanford Site in Washington are that they will be separated into high-level waste and low-activity waste fractions that will be vitrified separately. Formulating optimized glass compositions that maximize the waste loading in glass is critical for successful and economical treatment and immobilization of these nuclear wastes. Glass property-composition models have been developed and applied to formulate glass compositions for various objectives for the past several decades. Property models with associated uncertainties combined with composition and property constraints have been used to develop preliminary glass formulation algorithms designed for vitrification process control and waste-form qualification at the planned waste vitrification plant. This paper provides an overview of the current status of glass property-composition models, constraints applicable to Hanford waste vitrification, and glass formulation approaches that have been developed for vitrification of hazardous and highly radioactive wastes stored at the Hanford Site.

Behavior of technetium in nuclear waste vitrification processes

Nearly 100 tests were performed with prototypical melters and off-gas system components to investigate the extents to which technetium is incorporated into the glass melt, partitioned to the off-gas stream, and captured by the off-gas treatment system components during waste vitrification. The tests employed several simulants, spiked with 99mTc and Re (a potential surrogate), of the low activity waste separated from nuclear wastes in storage in the Hanford tanks, which is planned for immobilization in borosilicate glass. Single-pass technetium retention averaged about 35 % and increased significantly with recycle of the off-gas treatment fluids. The fraction escaping the recycle loop was very small.

Technetium Getters to Improve Cast Stone Performance

ABSTRACTA cementitious waste form known as Cast Stone is the baseline waste form for solidification of aqueous secondary wastes, including Hanford Tank Waste Treatment and Immobilization Plant (WTP) secondary liquid effluents to be treated and solidified at the Hanford Site Effluent Treatment Facility. Cast Stone is also being evaluated as a possible supplemental immobilization technology to provide the necessary low activity waste (LAW) treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. Two radionuclides of particular concern in these waste streams are technetium-99 (99Tc) and iodine-129 (129I). These radioactive tank waste components are predicted to contribute the most risk to groundwater – the most probable pathway for future environmental impacts associated with the cleanup of the Hanford site. A recent environmental assessment of Cast Stone performance, which assumes a diffusion controlled release of contaminants from the waste form, calculates groundwater in excess of the allowable maximum permissible concentrations for both contaminants. There is, therefore, a need and an opportunity to improve the retention of both 99Tc and 129I in Cast Stone. One method to improve the performance of Cast Stone is through the addition of “getters” that selectively sequester Tc and I, therefore reducing their diffusion out of Cast Stone. In this paper, we present results of Tc and I removal from solution with various getters. Batch sorption experiments were conducted with deionized water (DIW) and a highly caustic LAW simulant with a 7.8 M average Na concentration. In general, the data show that the selected getters are effective in DIW but their performance is compromised when experiments are performed with the 7.8 M Na Ave LAW simulant. The diminished performance in the LAW simulant may be due to competition with Cr present in the 7.8 M Na Ave LAW simulant and to a pH effect that may create a negatively charged surface that can repel negatively charged species.