Tritium waste control: April--June 1978. [Catalytic exchange detritiation; liquid waste decontamination; fixation in polymer impregnated concrete; management of high specific activity tritiated wastes]
The Combined Electrolysis Catalytic Exchange system was operated to experimentally determine mass transfer coefficients and to test the process controller. Values for H/sub OG/ and K/sub tilde y/a were obtained at three separate molar flow ratios (tilde L/tilde G). Replicate values of K/sub tilde y/a from additional runs agreed with initial results to within 16%. Two process controller tests were completed that demonstrated the reliability of the system hardware and the feasibility of the digital controller software. The feasibility of using a xenon flashlamp source in the uv photodissociation step of the two-photon water-hydrogen laser isotope separation (LIS) process has been demonstrated with H/sub 2/O/D/sub 2/ and D/sub 2/O/H/sub 2/ photocatalyzed exchange experiments. A nearly 10 : 1 isotopic selectivity between the photodissociation of ground state H/sub 2/O and D/sub 2/O was observed with an unfiltered xenon flashlamp source. The effectiveness of the hydrogen scavenger system was also demonstrated in these experiments. Tests continued on samples of cement and cement-plaster mixtures which were injected with tritiated water, cured, and then impregnated with catalyzed styrene monomer. After polymerization the samples were put into uncontaminated water and the tritium concentration was monitored. No significant differences were noted except in two cases when the polyethylene bottle had been removed, which resulted in 35 times more tritium being released into the surrounding water. The samples still in the polyethylene bottles have released an average of 2.3 Ci to the water. The tritium release study of actual burial packages is continuing. Two additional drums containing octane waste were added to the study, and now all types of liquid waste packaged are represented in the test. The average fractional release from three packages containing oil or water waste is 5 x 10/sup -7/ after 180 weeks.
- Research Organization:
- Mound Plant (MOUND), Miamisburg, OH (United States)
- DOE Contract Number:
- EY-76-C-04-0053
- OSTI ID:
- 6644265
- Report Number(s):
- MLM-2542
- Country of Publication:
- United States
- Language:
- English
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Tritium waste control: July--September 1978. [Catalytic exchange detritiation; fixation of aqueous tritiated waste in polymer-impregnated concrete; gas generation by self-radiolysis of polymer-impregnated concrete]
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Related Subjects
RADIOACTIVE WASTE PROCESSING
RESEARCH PROGRAMS
TRITIUM
CEMENTS
CONCRETES
ELECTROLYSIS
FEASIBILITY STUDIES
HIGH-LEVEL RADIOACTIVE WASTES
LIQUID WASTES
MASS TRANSFER
POLYMERS
BETA DECAY RADIOISOTOPES
BETA-MINUS DECAY RADIOISOTOPES
BUILDING MATERIALS
HYDROGEN ISOTOPES
ISOTOPES
LIGHT NUCLEI
LYSIS
MANAGEMENT
MATERIALS
NUCLEI
ODD-EVEN NUCLEI
PROCESSING
RADIOACTIVE MATERIALS
RADIOACTIVE WASTES
RADIOISOTOPES
WASTE MANAGEMENT
WASTE PROCESSING
WASTES
YEARS LIVING RADIOISOTOPES
052001* - Nuclear Fuels- Waste Processing