The leaching mechanisms of simulated low-level radioactive waste forms are being determined as support for development of an accelerated leach test. Two approaches are being used: (1) comparisons of leaching data with results of a model that describes diffusion from a finite cylinder, and (2) observation of the leaching process at temperatures between 20°C and 65°C. To provide results that can be used for modeling, leaching at elevated temperatures must change neither the leachins mechanism nor the structural controls of leaching such as the porosity. Releases of lj "Cs, Sr, calcium, sodium and potassium from Portland cement containing sodium sulfate. as a simulated evaporator sludge, have been determined under a variety of experimental conditions. Data from the leach tests were compared to model results for diffusion from the finite cylinder. While most leaching appears to be diffusion controlled, notable exceptions occur. For all samples activation energies ranging between 6 and 11 tCcal/mole have been calculated from the relationship of the effective diffusion coefficient to increasing temperature, close to the expected value of 5 Kcal/mole for diffusion.

No abstract

This report is a step-by-step guide for the Accelerated Leach Test (ALT) Computer Programdevelopedto accompanya new leach test for solidifiedwaste forms. The program is designedto be used as a toolfor performingthe calculations necessaryto analyze leachtest data, a modelingprogramto determine if diffusionis the operatingleachingmechanism(and, if not, to indicateother possiblemechanisms),and a meansto make extrapolationsusingthe diffusionmodels. The ALT program containsfour mathematicalmodelsthat can be used to representthe data. I _llr 12 SCREEN lA Use the cursor keys to highlight your choice, then press enter to choose it.

LIST OF FIGURES Figure No. Page 3.1 Compositional phase diagram for the solidification of aqueous waste containing 50 wt% or less ^SQ, with portland type I cement 8 3.2 Ternary compositional phase diagram for the solidification of incinerator ash in portland type I cement 8 3.3 Sodium ion concentrations released from vinyl ester-styrene/sodiurn sulfate waste forms after 3 day immersion in deionized water at 20°C versus waste loading 10 4.1 Cumulative fraction releases from portland cement for Sr-85 and Cs-137. No release of Co-60 was detected 21 4.2 Cumulative fraction releases for Cs-137 from portland cement. Final release was 0.817 of total activity in the waste form 23 4.3 Rate of Cs-137 release from portland cement. Note the change of rate between 10 and 20 days 23 4.4 Cumulative fraction release for portland cement samples. Data only goes out to 144 days because of radioactive decay of the i sotope 24 4.5 Release rates of Sr-85 from portland cement 24 4.6 Cumulative fraction releases of Cs-137, Sr-85, Co-60 from a vinyl ester-styrene emulsion waste form 25 4.7 Cumulative fraction releases of Cs-137 from triplicate VES emulsion waste forms. Note the scale used on the y-axis makes a small difference seem large « 26 4.8 Release rates for Cs-137 from triplicate VES emulsion waste forms. The scatter at the lower right of the plot indicates larger counting errors due to low release rates 26 4.9 Cumulative fraction releases for Sr-85 from VES emulsion waste forms. These data extend only to 144 days because radioactive decay reduced the count rate to detection limit activities 27 4.10 Release rates for Sr-85 for triplicate VES emulsion samples 27 4.11 Cumulative fraction releases for Co-60 from triplicate vinyl ester-styrene emulsion samples 28-viii-LIST OF FIGURES (cont.) Figure No. Page 4.12 Release rates of Co-60 from triplicate vinyl ester-styrene samples. Scatter at the bottom right is due to relatively large counting errors resulting from low release rates 29 4.13 Cumulative fraction releases for Cs-137, Co-60 and Sr-85 from bitumen. The isotopes from this material leach at different rates, unlike those in VES emulsion samples 28 4.14 Cumulative fraction releases of Cs-137 from triplicate bitumen waste forms. Note the scale of the y-axis indicating that intersample variability is quite small 31 4.15 Release rates of Cs-137 from triplicate bitumen samples 31 4.16 Cumulative fraction releases of Sr-85 for triplicate bitumen sampl es 32 4.17 Release rates of Sr-85 for triplicate bitumen samples 32 4.18 Cumulative fraction releases of Co-60 from triplicate bitumen waste forms. Note the scale of the y-axis 33 4.19 Release rates of Co-60 from triplicate bitumen samples. The scatter at the bottom right is due to large counting errors caused by low release rates 33 4.20 Cumulative fraction releases of Cs-137 from cement, VES emulsion and bitumen samples. While VES and bitumen have similar low fraction releases that of cement is 82% of the total original activity 34 4.21 Plot of conductance (umhos/cm) versus alkalinity...

This report presents the analytical results for tritium content of soil cores taken at the Barnwell, South Carolina, disposal site, field measurements at Barnwell, concentrations of free chelating agents in selected trench waters, and the analyses of water samples collected at the Maxey Flats, Kentucky, disposal site. Tritium contents in soil cores taken below the trenches show a decrease in tritium with depth to a minimum value at approximately ten meters, followed by an increase below this depth. This deeper maximum probably represents the downward movement of the previous years seasonal maxima for water infiltration into the trenches. This amount of downward migration from the trench bottom is approximately what would be expected based on the hydraulic conductivity of these sediments. Field measurements of trench waters at the Barnwell, South Carolina, disposal site indicate that the waters are chemically oxidizing regimes relative to those at Maxey Flats and West Valley. Analyses were performed to determine the amounts of free chelating agents DTPA, EDTA, and NTA in selected trenches at the Maxey Flats, West Valley, Barnwell, and Sheffield, disposal sites. Amounts of free chelating agents were generally below lj.ig/g, with one samp}e as high as 28••llg/g. No drastic changes in trench water compositions were observed relative to previous sampling at Maxey Flats. The experimental interceptor trenches contain detectable amounts of strontium and plutonium. Tritium contents vary from typical disposal trench levels (E7-E8 pCi/L) in trench IT-2E, downward four oders of magnitude in trench IT-5 in a decreasing trend along the line of experimental trenches.. .