Properties of radioactive wastes and waste containers. Progress report No. 5, April--June 1977

• Main Authors: Colombo, P., Neilson, R.M. Jr.
• Format: Report
• Language: English
• Published: Brookhaven National Laboratory 1977
• Subjects: Research Programs; Building Materials; Urea; Aldehydes; Shielding; Organic Compounds; Processing; Radioactive Wastes; Bitumens; Gamma Radiation; Chemical Reactions; Cleaning; Formaldehyde; Waste Processing; Radiations; Thermodynamic Properties; Specific Heat; Wastes 052000* > Nuclear Fuels > Waste Management; Polymers; Separation Processes; Other Organic Compounds; Phase Transformations; Materials; Reactor Materials; Portland Cement; 12 Management Of Radioactive And Non-Radioactive Wastes From Nuclear Facilities; Compression Strength; Solidification; Biodegradation; Tar; Radioactivity; Mechanical Properties; Ionizing Radiations; Physical Properties; Boric Acid; Dissolution; Inorganic Acids; Matrix Materials; Waste Management; Solid Wastes; Thermal Conductivity; Leaching; Management; Cements; Radioactive Materials; Decontamination; Carbonic Acid Derivatives; Radioactive Waste Processing; Amides; Organic Nitrogen Compounds; Hydrogen Compounds; Decomposition; Electromagnetic Radiation; Carbonic acid
• Online Access: https://doi.org/10.2172/6708861; https://digital.library.unt.edu/ark:/67531/metadc1195633/

Portland type II cement decontamination factors for free standing water were determined to be 1, 11, and 200 for /sup 137/Cs, /sup 85/Sr, and /sup 60/Co respectively, indicating ion exchange processes occurring within the cement matrix. The quantity of organic carbon that could be removed from solidification agent matrix materials on leaching in distilled water was measured. No organic carbon was present in the portland type II cement leachant water after ten days while 3.4 ppM, 34.2 ppM, and 9500 ppM organic carbon were present respectively in the bitumen, Dow polymer and urea-formaldehyde leachant water (300 ml). The compression strengths of portland type II cement and urea-formaldehyde waste forms containing various simulated wastes were measured. The effect of the waste/cement ratio on the compression strength of portland type II cement simulated waste forms was also studied. The gamma radiation shielding characteristics of portland cement, urea-formaldehyde, and bitumen waste forms were studied in reference to the maximum incorporated activity permissible in 55 gallon waste packages for transportation in ''non-exclusive use'' vehicles under 49CFR173. Thermal conductivity, specific heat, and thermal diffusivity values were obtained for portland cement, urea-formaldehyde, bitumen, and Dow polymer waste forms. Solidification verification studies were initiated for the Dow polymer at waste/binder ratios representative of those proposed for use by the vendor. Successful solidifications were obtained for all waste types tested with the exception of 12 wt.% boric acid. The 12 wt.% boric acid waste was successfully solidified after a proprietary pretreatment of the waste stream.