Properties of radioactive wastes and waste containers. Quarterly progress report, January--March 1977

Technical Report · Mon Aug 01 04:00:00 EDT 1977

DOI:https://doi.org/10.2172/6567177· OSTI ID:6567177

Colombo, P; Neilson, Jr, R M

The densities of the simulated wastes used in this program were determined and factors reported to allow conversion of formulation data for urea-formaldehyde (UF) waste forms from weight ratios to volume ratios of waste to UF. Simulated Solka-Floc wastes were solidified with urea-formaldehyde but areas of unsolidified Solka-Floc were found in the final waste forms. The difficulty in mixing dewatered Solka-Floc waste with UF is presumed to be responsible. Attempts to solidify alkaline regenerative wastes with urea-formaldehyde were not successful. Increasing the waste/UF volume ratio to 3.0 produced acceptable solids with boric acid and decontamination waste, but not with chemical regenerative wastes. The use of masonry cement (one part Portland cement to one part anhydrous lime by weight) for the solidification of boric acid concentrate waste was investigated. Free standing solids were formed at waste/cement weight ratios of 0.6 and 1.0 but not at 1.5 or 2.0. Masonry cement was found to be superior to Portland cement for the solidification of boric acid wastes. The weight loss with temperature of urea-formaldehyde waste forms containing various simulated wastes was measured over the range of 25 to 500/sup 0/C. Weight losses due to the evaporation of water, decomposition of UF and decomposition of the wastes were observed. Static leaching experiments were conducted to determine the rate of release of cesium-137 and strontium-85 from urea-formaldehyde specimens and bead and powdered cation exchange resins. Results for the leaching of these materials in salt water, tap water, and distilled water are reported. Hydrostatic testing of modified DOT 17H open head 55 gallon drums was performed. These drums met the required DOT specifications and ultimately failed at an average pressure of 44 psig by separation of the drum head from the drum body at the gasket.

• Research Organization: Brookhaven National Lab., Upton, NY (USA)
• DOE Contract Number: EY-76-C-02-0016
• OSTI ID: 6567177
• Report Number(s): BNL-NUREG-50692
• Country of Publication: United States
• Language: English

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12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES
ALDEHYDES
ALKALI METAL ISOTOPES
ALKALINE EARTH ISOTOPES
AMIDES
BETA DECAY RADIOISOTOPES
BETA-MINUS DECAY RADIOISOTOPES
BUILDING MATERIALS
CARBONIC ACID DERIVATIVES
CEMENTS
CESIUM 137
CESIUM ISOTOPES
CONTAINERS
DAYS LIVING RADIOISOTOPES
DENSITY
DISSOLUTION
ELECTRON CAPTURE RADIOISOTOPES
EVEN-ODD NUCLEI
FORMALDEHYDE
HOURS LIVING RADIOISOTOPES
INTERMEDIATE MASS NUCLEI
ISOMERIC TRANSITION ISOTOPES
ISOTOPES
LEACHING
MANAGEMENT
MATERIALS
NUCLEI
ODD-EVEN NUCLEI
ORGANIC COMPOUNDS
ORGANIC NITROGEN COMPOUNDS
PERFORMANCE TESTING
PHASE TRANSFORMATIONS
PHYSICAL PROPERTIES
PROCESSING
RADIOACTIVE MATERIALS
RADIOACTIVE WASTE PROCESSING
RADIOACTIVE WASTES
RADIOISOTOPES
RESEARCH PROGRAMS
SEPARATION PROCESSES
SOLIDIFICATION
STRONTIUM 85
STRONTIUM ISOTOPES
TESTING
UREA
WASTE MANAGEMENT
WASTE PROCESSING
WASTES
YEARS LIVING RADIOISOTOPES