Cementitious Waste Forms Research Articles

Leaching from cementitious waste forms in belowground vaults

Leaching from cementitious waste forms in belowground vaults

Geochemical engineering of low level radioactive waste in cementitious environments

Significantly reduced release of radionuclides from low level radioactive waste disposal facilities may be achieved by passively engineering the geochemical environment in and immediately adjacent to shallow land burial sites. Mineral growth and adsorption onto cementitious waste forms, casks, and backfills is calculated to minimize the transport of 234,238U, 241Pu, 241Am, 232Th, 59,63Ni and 14C. Reductive adsorption of Tc to Fe(II)-rich minerals may be engineered by emplacing a fine-grained mafic (basalt, serpentinite, scrap iron) blanket of aggregate below and around the LLRW repository. 90Sr and 137Cs transport into the biosphere may also be retarded due to adsorption on cementitious material, through scavenging by the Fe-oxyhydroxide weathering products of the mafic blanket, and by dilution.

Equilibrium concentration and diffusion controlled leaching model for cement-based waste forms

A predictive model for nuclide leaching from cementitious waste form is developed based on diffusion, the sorption equilibrium concept in the waste form, and continuity boundary conditions of concentration and concentration gradient on the waste form surface. A concept to determine the equilibrium concentrations of radionuclides in pore water of the waste form is presented. Nuclide release unrelated to diffusive transport mechanism is reflected on the model as a wash-off fraction which depends on the surface conditions of the waste form. The double region model proposed in this paper gives much lower nuclide release rates than the single region model. This leaching model is calibrated by comparison with the results of a leaching experiment performed at JAERI.

Chromium Stabilization Chemistry of Paint Removal Wastes in Portland Cement and Blast Furnace Slag

The use of cement based systems for solidification and stabilization of hazardous wastes has been proposed. The stabilization of Cr contaminated paint removal wastes in ordinary Portland cement and in a Portland cement and blast furnace slag matrix was investigated. A loading by volume of 75% waste and 25% cement (or cement + slag) was used. The expression of pore solution was utilized to determine the chemical environment encountered by the waste species in the cement matrix. The highly alkaline conditions of ordinary Portland cement determined the stability of the metal species, with Cr being highly soluble. The replacement of 25% of the Portland cement by blast furnace slag was found to decrease the [OH-] of the pore solution resulting in a decrease of the Cr concentration. For cement wastes forms hydrated for 28 days, the Cr concentration decreased in the expressed pore solution. During the TCLP tests the cement waste form and extraction solution were found to react, changing the chemistry of the extraction solution. The expression of pore solution was found to give a direct measure of the chemistry of the waste species in the cement matrix. This avoids the reaction of the TCLP extraction solution with the cement matrix which changes the solubility of the hazardous metals.

Fixation of Copper Ferrocyanide Cake in the urea-formaldehyde matrix

Fixation of Copper Ferrocyanide Cake (CFC), which is generated from the chemical treatment plant of low-level radioactive liquid effluent in a urea-formaldehyde (UF) matrix, was carried out. Boric acid was used as a catalyst for the UF polymerization. The physical properties such as compressive strength and diffusion coefficient were evaluated for UF waste forms and were compared with those of Ordinary Portland Cement (OPC) waste forms. While the compressive strengths of UF and OPC waste forms were comparable, the former showed diffusion coefficients about two to three orders lower. Almost all the cesium was released from the OPC waste form that contained less than 18% of CFC by weight in the final waste form. The release of cesium from the OPC waste forms was high due to the presence of high alkalinity and calcium ions in the cement. To find out the temperature rise in the preparation of UF waste form, blocks having a size of about 5 1 were prepared; it was observed that there was no increase in peak center temperatures during the curing process.

Hot Isostatic Press (Hip) Vitrification of Radwaste Concretes

AbstractProperly prepared cementitious waste forms can be hot-isostatically-pressed into materials that exhibit performance equivalent to typical radwaste-type glasses. The HIPing conditions (temperature/pressure) required to “vitrify” these concretes are quite mild and therefore consistent with both safety and good productivity. This paper describes both the process and its products with reference to potential application to Idaho Chemical Processing Plant (ICPP) reprocessing wastes.

Leaching from Cementitious Waste Forms in Belowground Vaults

Solidification and/or stabilization with cementitious materials prior to burial is one option for disposal of liquid hazardous and radioactive wastes. A common design for disposal of cementitious waste forms is to pour the material into large belowground vaults. The leaching performance of partially degraded monolithic vaults is examined quantitatively for facilities located in humid to semiarid climates. Development of perched water on the vault roof leading to fracture flow through the structure is predicted for a wide range of climate and design conditions. Leaching controlled by diffusion in matrix blocks out to fractures is examined parametrically in relation to water flux rate and crack spacing. Depending upon the parameters examined, release rate may be controlled by water flux rate (advection controlled) or diffusion. Under some circumstances, contaminant release rates and exit concentrations are predicted to be inversely related. In this situation, minimization of release does not result in the l...

Generation Mechanism of Hydrogen Gas from Hardened Cement Paste by γ-Irradiation

AbstractHydrogen gas is generated from cementitious waste forms by radiolysis of water. In the case of low level radioactive waste, gas yields have been confirmed to be sufficiently low by irradiation experiments. However, studies have suggested that the hydrogen generation rate in cementitious waste forms is larger than the rate calculated from the g-value (H2 yields for 100eV absorbed). In this paper, the factors that increase the gas generation were investigated quantitatively. Two factors were identified, the effect of an organic diethylene glycol which reacts with hydrogen radicals to produce hydrogen, and the effect of electrons generated in the cementitious matrix which decompose water to hydrogen. The hydrogen generation rate was confirmed to drop less than the rate calculated from the g-value when these factors were eliminated.

Carbon-14 behaviour in a cement-dominated environment: Implications for spent candu resin waste disposal

Cement based waste forms and concrete engineered barriers are expected to play a key role in providing 14 C waste containment and control 14 C migration for time periods commensurate with its hazardous life of about 50,000 years. The main thrust of this study was, therefore, to evaluate the performance of cement based waste forms with regard to 14 C containment. Of particular importance are the geochemical processes controlling 14 C solubility and release under anticipated cement dominated low and intermediate level waste repository conditions. Immobilization of carbonate-form exchange resin in grout involves transfer of sorbed 14 CO 3 2− ions, through exchange for hydroxyl ions from the grout slurry, followed by localized precipitation of solid calcium carbonate at the cement/resin interface in the grout matrix. Thus almost a total stoichiometric transfer of labile carbonate, originally present on the spent resin, to a solid carbonate phase takes place upon immobilization of carbonate-form resin in a cementitious matrix. Carbon-14 release behaviour can be attributed to the dissolution characteristics and solubility of calcite present in the cement based waste form. The groundwater flow regime can exert a pronounced effect both on the near-field chemistry and the leaching behaviour of 14 C. For a cement dominated repository, at relatively low-flow or stagnant groundwater conditions, the alkaline near-field chemical environment inhibits the release of 14 C from the cemented waste form. Under high flow conditions, the near-field environment is characterized by relatively neutral pH conditions which promote calcite dissolution, thus resulting in 14 C release from the waste form. These findings point to the importance of selecting a suitable site for 14 C waste disposal with a flow regime so that local equilibrium prevails and 14 C release becomes solubility-limited and coupled to generation and maintenance of alkaline conditions for a long period of time.

Desorption Kinetics of Radionuclides Fixed in Cement Matrix

Leach characteristics of some typical nuclides such as cesium, strontium, cobalt, I[sup [minus]], and CO[sup [minus]2][sub 3] from ordinary portland cement waste forms have been studied using the ISO test method and radiotracers of the respective nuclides, i.e., [sup 134]Cs, [sup 85]Sr, [sup 60]Co, [sup 131]I, and [sup 14]C. The leach studies suggest a rapid release of radioactivity in the beginning (fast component) followed by slow release for long periods of time (slow component). A mathematical model has been simulated to describe the leaching kinetics of these nuclides from the cement matrix. The effective diffusion coefficient D[sub [ell]] is computed from the two componental diffusion coefficients, and the retardation factor ([alpha]) for a nuclide is evaluated from a knowledge of the radioactivity distribution in the aqueous and solid phase at equilibrium. The product [alpha]D[sub [ell]] for all the nuclides studied has been found to be approximately constant and is equal to the intrinsic diffusion coefficient in the cement matrix. The net fractional release of different radionuclides from cement waste form showed a decreasing pattern, i.e., [sup 134]Cs > [sup 131]I > [sup 85]Sr > [sup 14]Cr > [sup 60]Co indicating the largest diffusion coefficient for cesium as 10[sup [minus]2] cm[supmore » 2]/day and the least for [sup 14]C as 3 [times] 10[sup [minus]8] cm[sup 2]/day.« less

Ancient Analogues Concerning Stability and Durability of Cementitious Wasteform

ABSTRACTThe history of cemenütious materials goes back to ancient times. The Greeks and Romans used calcined limestone and later developed pozzolanic cement by grinding together lime and volcanic ash called “pozzolan” which was first found near Port Pozzuoli, Italy. The ancient Chinese used lime-pozzolanic mixes to build the Great Wall. The ancient Egyptians used calcined impure gypsum to build the Great Pyramid of Cheops. The extraordinary stability and durability of these materials has impressed us, when so much dramatically damaged infrastructure restored by using modern portland cement now requires rebuilding. Stability and durability of cementitious materials have attracted intensive research interest and contractors’ concerns, as does immobilization of radioactive and hazardous industrial waste in cementitious materials. Nuclear waste pollution of the environment and an acceptable solution for waste management and disposal constitute among the most important public concerns. The analogy of ancient cementitious materials to modern Portland cement could give us some clues to study their stability and durability. This present study examines selected results of studies of ancient building materials from France, Italy, China, and Egypt, combined with knowledge obtained from the behavior of modern portland cement to evaluate the potential for stability and durability of such materials in nuclear waste forms.

Improved cement barriers applied in nuclear wastes

Laboratory scale experiments have been carried out to compare the rate of release of radionuclides from cemented waste forms with unprocessed powder waste in preformed cement containers. Attempts to improve the properties of cement matrices used as a barrier against the release of radionuclides from waste materials focus on chemical treatment and polymer impregnated cement (PIC). The results obtained show that the incremental release of cesium per day per cm 2 of exposed surface area was decreased for processed cement-waste forms (chemically treated and / or polymer impregnated) compared with unprocessed cement-waste forms. Further improvement in the incremental release was achieved using unprocessed powder wastes in cement containers. Chemical treatment of waste material improved the cumulative percent release of cement — waste forms by a factor of four for cesium and by ten for cobalt.

Leaching behavior of carbon-14 contained in portland cement

MCC-1 static leaching experiments were carried out for a cementitious waste form in distilled water for up to 64 days at 5°C and 20°C in order to examine the leaching behavior of carbon-14. The complicated leaching behavior of carbon-14, meaning that the leached carbon-14 activity did not increase with (time) 0.5, was attributable to the precipitation of calcite and the formation of colloidal particles in leachates, which were mainly dependent on the pH value and calcium concentration of leachate. The normalized elemental mass loss of carbon-14 was about 7.5 × 10 −4 g/cm 2 at 20°C for 64 days, which was lower than those of cement constitute elements such as calcium, sodium and aluminum. Especially, the leach rate of aqueous carbon-14 was lower than that of carbon-14 in the suspended leachate by a factor of about 10.

Reduction of Cesium Leaching Ratio from Cementitious Resin Forms Using Natural Active Silica.

Cesium adsorption behavior of active silica, which is a natural acid clay composed of cristobalite and quartz, was evaluated for its applicability as Cs adsorbent to be added to cementitious waste forms containing spent ion exchange resin. Since active silica carried the Cs exchangeable silanol group (—SiOH) originally, the Cs distribution coefficient was remarkably high (<104). It increased in saturated Ca(OH)2 solution, simulating the cement paste, due to formation of new silanol groups. With its addition to the cementitious forms with 134Cs adsorbed ion exchange resin solidified by slag cement, the Cs leaching ratio was reduced to below 1/10 that without active silica.

Evaluation of the Retention Capability of Backfill Materials in the Presence of Cement Waste Form

ABSTRACTThe present study focused on the retention capability of the different local backfill materials and on horizontal and vertical radionuclide migration in simulated repository conditions of a saturated static humid environment, using single or combined components of the near-field. The results obtained from semi-field experiment show that no migration of cesium radionuclide was detected outside the backfill zone within the time interval of the experiment. This reflects the possible efficiency, of the backfill materials used, for the confinement of radioactivity to the disposal site. On the other hand laboratory experiments show the effect of simulated repository condition on the sorption and desorption properties of backfill materials. It is clear from the results obtained that the presence of cement waste forms in equilibrium with underground water affect the retention capability of the backfill materials. The motivation of the work was a desire to provide a basis for minimizing radioactive waste processing by improving nonradioactive engineering barriers.

Cement-based engineered barriers for carbon-14 isolation

This paper presents a review of near-field investigations relevant to the use of cement-based engineered barriers for the long-term isolation of Ontario Hydro's 14C reactor operating waste. Specifically, the research program was designed to evaluate the effectiveness of cemented waste form and cementitious grout backfill for radiocarbon waste isolation in a repository setting. Carbonation of grout results in a calcitic end-product which has the potential to immobilize radio-carbon. 14C waste form evaluation studies indicate almost a total stoichiometric transfer of labile carbonate to the solid carbonate phase upon immobilization of carbonate-form resin in a cementitious matrix. The observed 14C leaching behaviour can be attributed to the dissolution characteristics of calcite in the waste form. The chemical buffering of the near-field groundwater environment by the cementitious matrix will ensure low solubility of calcite, thereby contributing to the retention of 14C in the waste form. Although the carbonation capacities of the backfill materials studied range from 16% to 19% of the grout mass, this high degree of carbonation was not attained for solid grout samples. Carbonation of grout not only contributes to the effectiveness of the engineered barrier as a geochemical sink for 14C but also causes a significant reduction in the permeability, resulting in diffusion-controlled mass transfer.

Effect of Porous Surface Layer on Leaching Rate of Radionuclide from Cementitious Waste Form

Leach tests based on IAEA methods were carried out using three types of cementitious specimens containing 134Cs and 125I and an evaluation method of apparent diffusion coefficients from the leaching curves was discussed. Since the gradient of the leaching curve, i.e. the relationship between the leaching ratio and the square root of time, changes during the leach tests, it has been difficult to estimate the diffusion coefficient. The difference between the leaching rate in the early period of the leach test and that in the later period was attributed to the porous microstructure of the surface layer in the cementitious specimens. The duplicate layer model, which expressed differences in the properties between the surface layer and an inner region by different diffusion coefficients, was proposed. Its numerical solutions had good agreement with the experimental leaching curves.

Effect of Porous Surface Layer on Leaching Rate of Radionuclide from Cementitious Waste Form.

Leach tests based on IAEA methods were carried out using three types of cementitious specimens containing 134Cs and 125I and an evaluation method of apparent diffusion coefficients from the leaching curves was discussed. Since the gradient of the leaching curve, i.e. the relationship between the leaching ratio and the square root of time, changes during the leach tests, it has been difficult to estimate the diffusion coefficient. The difference between the leaching rate in the early period of the leach test and that in the later period was attributed to the porous microstructure of the surface layer in the cementitious specimens. The duplicate layer model, which expressed differences in the properties between the surface layer and an inner region by different diffusion coefficients, was proposed. Its numerical solutions had good agreement with the experimental leaching curves.

Leach rate expressions for performance assessment of solidified, low-level radioactive waste

Exact, analytical expressions for the diffusive leach rate of radioactive species from low-level, cylindrical, and rectangular-shaped wasteforms are obtained based on typical assumptions about the wasteform as are codified in the American National Standards Institute/American Nuclear Society ANSI/ANS 16.1 leachability standard of 1986. The new expressions are entirely analytical and model both the axial and the radial (for the cylindrical geometry) leach rate as a function of time, wasteform dimensions, diffusion coefficient of the species under study, and decay constant. Improved leach rate expressions for implementation in performance assessment codes of disposal sites are also provided along with an analysis of the associated errors. The numerical analysis utilizes parameter values applicable to strontium and faster-leaching nuclides from full-size, cement wasteforms under disposal conditions. With the decay constant set equal to zero, all the proposed expressions also apply to stabilized hazardous wastes and their disposal sites.

Improved expressions for modeling diffusive, fractional cumulative leaching from finite-size waste forms

The classical expressions for modeling diffusive, fractional cumulative leach data from finite-size waste forms are in the form of open series of slow convergence and difficult error control. Alternatively, closed form expressions are obtained which are entirely analytical. Simple analytical formulae for the error associated with the proposed expressions are also provided. The new expressions for reporting fractional cumulative leaching versus time are easy to implement numerically and are shown to be adequate for the modeling of typical laboratory leach data from cementitious waste forms.