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Development of ultrafiltration and inorganic adsorbents for reducing volumes of low-level and intermediate-level liquid waste: April--June 1977

Description: Ultrafication (UF) membranes have demonstrated 90 to 98% rejection of gross alpha in laboratory tests. In the treatment of laundry wastes, rejection of activity ranged from 98 to 99.9% gross alpha. The pilot UF system was installed and started up. Flux decline curves and volume reduction performance were determined. Volume reductions of 210 : 1 were achieved at flux rates of 1.1 gal/min (system is rated at 2 to 3 gal/min, 90% recovery) at activity rejection of 99.94% gross alpha. Adsorbent studies demonstrated capacities in excess of 10/sup 9/ dis/min/g for uranium-233 and in excess of 10/sup 8/ dis/min/g for plutonium-238. Construction and start-up of the Engineering Test Facility has been completed.
Date: November 14, 1977
Creator: Koenst, J.W.; Herald, W.R. & Roberts, R.C.
Partner: UNT Libraries Government Documents Department

Separation of tritium from gaseous and aqueous effluent systems

Description: Three processes are discussed for separating tritium from gaseous and aqueous effluent systems: separation in the gas phase using Pd-25 wt percent Ag alloy diffusion membranes; electrolytic separation in the aqueous phase using ''bipolar'' electrodes; and the countercurrent exchange of tritium-containing hydrogen gas with water on catalytic surfaces combined with separation by direct electrolysis. (LK)
Date: January 1, 1977
Creator: Kobisk, E.H.
Partner: UNT Libraries Government Documents Department

Decontamination of the Curium Source Fabrication Facility

Description: The Curium Source Fabrication Facility (CSFF) at Oak Ridge National Laboratory (ORNL) was decontaminated to acceptable contamination levels for maintenance activities, using standard decontamination techniques. Solid- and liquid-waste volumes were controlled to minimize discharge to the ORNL Waste Systems. This program required two years of decontamination effort at a total cost of $580K.
Date: January 1, 1982
Creator: Schaich, R.W.
Partner: UNT Libraries Government Documents Department

Accelerator transmutation of sup 129 I

Description: Iodine-129 is one of several long-lived reactor products that is being considered for transmutation by the Los Alamos Accelerator Transmutation of Waste (ATW) program. A reasonable rate of transmutation of 1291 is possible in this system because of the anticipated high neutron flux generated from the accelerator. This report summarizes previous papers dealing with the transmutation of 1291 where reactor technologies have been employed for neutron sources. The transmutation process is considered marginal under these conditions. Presented here are additional information concerning the final products that could be formed from the transmutation process in the ATW blanket. The transmutation scheme proposes the use of solid iodine as the target material and the escape of product xenon from the containers after van Dincklange (1981). Additional developmental plans are considered.
Date: January 1, 1992
Creator: Attrep, M. Jr.
Partner: UNT Libraries Government Documents Department

Tracer-level radioactive pilot-scale test of in situ vitrification technology for the stabilization of contaminated soil sites at ORNL

Description: This plan summarizes the activities to be performed during FY 1990 and FY 1991 for the tracer-level radioactive pilot-scale in situ vitrification (ISV) test. This test is the second step in evaluating ISV as a remedial action for the pits and trenches at Oak Ridge National Laboratory (ORNL). A previous test used nonradioactive tracers for cesium and strontium. This new test will again use a one-half-scale model of trench 7 and the pilot-scale ISV equipment of Pacific Northwest Laboratory (PNL). A small and precisely known amount of waste from a liquid waste disposal pit will be used for the test. An actually contaminated waste site cannot be used for this test because of the necessity to use an exactly known inventory of radionuclides so that a precise measurement of the volatilization of various constituents to the off-gas can be determined.
Date: November 1, 1991
Creator: Jacobs, G.K. & Spalding, B.P.
Partner: UNT Libraries Government Documents Department

Tritium waste control, October--December 1977. [Tritiated liquid waste decontamination; fixation of aqueous tritiated waste in polymer-impregnated concrete]

Description: The combined electrolysis-catalytic-exchange pilot scale system was operated on a 5 hr/day basis for 32 days without a major equipment failure. Modifications were made to the Nd:YAG laser that increased the output power by five. The increased beam quality and power of the laser allowed optical parametric oscillator (OPO) operation over a 2.9 to 4.14 ..mu..m tuning range. Background exchange was eliminated in the photolysis system and photolytic experiments were commenced. Photo dissociation experiments with H/sub 2/O/D/sub 2/ revealed that the xenon flashlamp should have sufficient spectral emittance in the less than or equal to 194 nm region to produce detectable amounts of photodissociation in vibrationally excited HDO. The short 10 cm long x 0.6 cm i.d. column was replaced by a 50-cm-long column of the same diameter. A thermometer was placed at the midpoint of the column to assist in the analysis of column operation. A new liquid level probe was installed. Tests of stage height were repeated with the longer column. Triplicate samples of cement, cement-plaster (1:1 ratio by weight), and cement-plaster (1:1 ratio by volume) were injected with 386 Ci of tritium, cured for five days, and then impregnated with catalyzed styrene monomer. After polymerization, the samples were put into a test program to measure the tritium release. When the polystyrene bottle was removed the tritium release increased fiftyfold. This corresponds to a fractional release of 2.8 x 10/sup -1/ after 31 weeks. The tritium release study of actual burial packages is continuing. Two additional drums have just been added to the study. These contain octane waste. The fractional release is 1 x 10/sup -5/ on a 5-yr old package, only 4.0 x 10/sup -7/ on a 4-yr old package, and 4.7 x 10/sup -9/ on the 2-yr old package.
Date: March 4, 1978
Partner: UNT Libraries Government Documents Department

Tilt-pour melt-caster for encapsulation of radioactive cesium

Description: Use of the tilt-pour melt-caster makes distinct improvements in the cesium encapsulation process. Compared to the vacuum castings system now in use, the tilt-pour equipment requires no reliance on heat-traced transfer lines, less sealing pressure for capsule filling, is less corrosive to capsules and is easier to repair. From the results of the extensive development program, it is concluded that the tilt-pour melt-caster can be operated to meet the cesium encapsulation production and maintenance requirements.
Date: January 1, 1977
Partner: UNT Libraries Government Documents Department

Development of ultrafiltration and inorganic adsorbents for reducing volumes of low-level and intermediate-level liquid waste: January--March 1978

Description: The exposure of noncellulosic ultrafiltration membranes to a radioactive environment simulating up to 24 months of exposure to a beta dose of 10 ..mu..Ci/cm/sup 3/ and a gamma dose of 10/sup -5/ ..mu..Ci/cm/sup 3/ did not show any conclusive evidence of membrane degradation. Viscosity measurements for control membranes and irradiated membranes indicate no changes in polymer molecular weight were caused by the radiation exposure. This, in turn, suggests no physical or mechanical degradation took place. A continuous run on the ultrafiltration pilot plant was completed. The run lasted 33.5 hr; during this period, the flux declined from 4 gal/min to 0.8 gal/min while rejection of alpha activity increased slightly. A total of 20,000 gal were processed through the ultrafiltration system. The small laboratory column tests were continued with uranium-233 and americium-241. Several new resins were being evaluated along with the same type of resin as used before with uranium-233 and plutonium-238. Tests were continued with the 2-in. Engineering Columns using ultrafiltration product spiked with uranium-233.
Date: April 27, 1978
Creator: Koenst, J.W.; Herald, W.R. & Roberts, R.C.
Partner: UNT Libraries Government Documents Department

Volume reduction system for solid and liquid TRU waste from the nuclear fuel cycle: October--December 1977

Description: A bench-scale incinerator for the investigation of unusual particulate and gaseous radioactive material in the incinerator off-gas was assembled and equipment checkout was begun. Modifications to a glovebox to be used for the demonstration phase of incinerator-ash immobilization are approximately 80% completed with demonstration to begin next quarter. The curing time for ash-cement pressed pellets was optimized by periodic crush-strength testing of open-air and underwater cured samples. Leach tests were performed at 70 and 100/sup 0/C to simulate normal and extreme storage conditions. Long-term leach tests were initiated on plutonium-238 doped ash-cement pressed pellets in distilled water at ambient temperature. Pressed pellets of bone char, ash, and sludge-ash in several mixtures were fabricated and tested to establish pressing and curing parameters. Leach tests were also performed on bone char-cement pressed pellets. Curing studies have been conducted on the pressed pellet matrix to define differences between curing in the open atmosphere and under water. Leachability studies have been initiated on the pressed pellet ash/cement matrix in distilled water at ambient, 70 and 100/sup 0/C. Also, leachability studies on doped plutonium-238 pressed ash/cement pellets and pressed bone char/cement pellets in distilled water at ambient temperature have been conducted. Compressive strengths of bone char/cement and sludge/cement pressed matrix have been investigated.
Date: April 21, 1978
Creator: Luthy, D. F. & Bond, W.H.
Partner: UNT Libraries Government Documents Department

Technology transfer: Ion exchange resins for Technetium-99 removal from X-705 raffinates

Description: An ion exchange process will be used at Portsmouth to remove Technetium-99 from uranium recovery waste solutions (raffinates). Subsequent treatment will then remove nitrates from the raffinates by a biodenitrification process prior to discharge to receiving streams to meet environmental standards for liquid wastes. Ion exchange process parameters affecting safe and efficient raffinate treatment have been examined in the laboratory, and results are described in this report. 4 refs., 3 figs., 6 tabs.
Date: December 3, 1982
Creator: Deacon, L.E. & Greiner, M.J.
Partner: UNT Libraries Government Documents Department

Solid state storage of radioactive krypton in a silica matrix

Description: The feasibility of loading a low density SiO/sub 2/ glass with krypton for storage of radioactive /sup 85/Kr has been demonstrated by studies using non-radioactive krypton. A 96% SiO/sub 2/ glass with 28% porosity was heated at an elevated pressure of Kr gas to a temperature of 850 to 900/sup 0/C and held at that temperature to sinter the glass-krypton composite to a density of about 2 g/cm/sup 3/. A krypton content of 30 cm/sup 3/ of Kr(STP)/cm/sup 3/ of glass has been demonstrated when loading pressures of 140 MPa are used. Krypton release rates from the glass are lower than reported for any other waste form considered currently. At 420/sup 0/C a diffusion parameter, D/r/sub 0//sup 2/, of 8.66 x 10/sup -13/ min/sup -1/ was determined which leads to a total release of 0.7% of the krypton in 10 years. Release rates increase moderately with increasing temperature up to 600/sup 0/C and increase rapidly above 600/sup 0/C. The lower loading pressures (about 40 MPa) may appear to yield a more favorable product from the point of view of krypton release than the high pressures. Advantages and disadvantages of the technique are given in the conclusions section.
Date: December 1, 1980
Creator: Tingey, G.L.; Lytle, J.M.; Gray, W.J. & Wheeler, K.R.
Partner: UNT Libraries Government Documents Department

Development of ultrafiltration and inorganic adsorbents for reducing volumes of low-level and intermediate-level liquid waste, April--June 1978

Description: A series of runs was performed in which waste processing facility influent was spiked with americium-241, neptunium-237, and uranium-233 and run through the ultrafiltration and reverse osmosis (RO) units. The results of these experiments show that the ultrafiltration membranes are ionic dependent, whereas the RO unit is not. Membrane irradiation studies have been started. Continuous run parameters are being verified through a series of experiments. The small laboratory column tests were continued this quarter on several adsorbents. Decontamination factors were calculated for these adsorbents in removing neptunium-237 and americium-241 from waste solutions. Tests were continued with the 2-in. Engineering Columns using ultrafiltration product spiked with uranium-233. A 6-in. diameter column was installed in the combined raffinate line from the three Engineering Columns. This ''mixed bed'' column will polish the waste solution that is returned to the waste processing facility tanks. A quality control program was started this quarter.
Date: July 19, 1978
Creator: Herald, W.R. & Roberts, R.C.
Partner: UNT Libraries Government Documents Department

Human factors engineering for the TERF (Tritium Emissions Reduction Facility) project. [Tritium Emissions Reduction Facility]

Description: The Tritium Emissions Reduction Facility (TERF) is being built by EG G Mound Applied Technologies to provide improved control of the tritium emissions from gas streams being processed. Mound handles tritium in connection with production, development, research, disassembly, recovery, and surveillance operations. During these operations, a small fraction of the tritium being processed escapes from its original containment. The objective of this report is to describe the human factors engineering as performed in connection with the design, construction, and testing of the TERF as required in DOE Order 6430.1A, section 1300-12. Human factors engineering has been involved at each step of the process and was considered during the preliminary research on tritium capture before selecting the specific process to be used. Human factors engineering was also considered in determining the requirements for the TERF and when the specific design work was initiated on the facility and the process equipment. Finally, human factors engineering was used to plan the specific acceptance tests that will be made during TERF installation and after its completion. These tests will verify the acceptability of the final system and its components. 16 refs., 8 figs.
Date: December 14, 1990
Creator: Hedley, W.H.; Adams, F.S. (EG and G Mound Applied Technologies, Miamisburg, OH (USA)) & Wells, J.E. (Lawrence Livermore National Lab., CA (USA))
Partner: UNT Libraries Government Documents Department

Characterization of off-gases from a small-scale, joule-heated ceramic melter for nuclear waste vitrification. [Ru, Cl, F, /sup 137/Cs]

Description: This paper confirmed with actual nuclear waste the thermodynamic predictions of the fate of some of the semivolatiles in off-gas. Ruthenium behaves erratically and it is postulated that it migrates as a finely divided solid, rather than as a volatile oxide. Provisions for handling these waste off-gasses will be incorporated in the design of facilities for vitrifying SRP waste.
Date: January 1, 1980
Creator: Woolsey, G.B. & Wilhite, E.L.
Partner: UNT Libraries Government Documents Department

Design and operation of a remotely operated plutonium waste size reduction and material handling process

Description: Noncombustible /sup 238/Pu and /sup 239/Pu waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant, and is being retrievably stored there. As part of the long-term plant to process the stored waste and current waste for permanent disposal, a remote size reduction and material handling process is being cold-tested at Savannah River Laboratory. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system has been completed. Fabrication and acceptance testing of the Telerobat, a robotically controlled manipulator has been completed. Testing is scheduled to begin in 3/86. Design features maximizing the ability to remotely maintain the equipment were incorporated. Complete cold-testing of the equipment is scheduled to be completed in 1987.
Date: January 1, 1986
Creator: Stewart, III, J A & Charlesworth, D L
Partner: UNT Libraries Government Documents Department

Design of zeolite ion-exchange columns for wastewater treatment

Description: Oak Ridge National Laboratory plans to use chabazite zeolites for decontamination of wastewater containing parts-per-billion levels of {sup 90}Sr and {sup 137}Cs. Treatability studies indicate that such zeolites can remove trace amounts of {sup 90}Sr and {sup 137}Cs from wastewater containing high concentrations of calcium and magnesium. These studies who that zeolite system efficiency is dependent on column design and operating conditions. Previous results with bench-scale, pilot-scale, and near-full-scale columns indicate that optimized design of full-scale columns could reduce the volume of spent solids generation by one-half. The data indicate that shortcut scale-up methods cannot be used to design columns to minimize secondary waste generation. Since the secondary waste generation rate is a primary influence on process cost effectiveness, a predictive mathematical model for column design is being developed. Equilibrium models and mass-transfer mechanisms are being experimentally determined for isothermal multicomponent ion exchange (Ca, Mg, Na, Cs, and Sr). Mathematical models of these data to determine the breakthrough curves for different column configurations and operating conditions will be used to optimize the final design of full-scale treatment plant. 32 refs., 6 figs., 3 tabs.
Date: January 1, 1991
Creator: Robinson, S.M.; Arnold, W.D. & Byers, C.H.
Partner: UNT Libraries Government Documents Department

Decontamination of low-level liquid waste at Oak Ridge National Laboratory using a scavenging-precipitation, ion exchange process

Description: For five and one-half years, liquid wastes at Oak Ridge National Laboratory (ORNL) which contained low levels of beta and gamma activity have been decontaminated by a scavenging-precipitation, ion exchange process (SP-IX). Some of the activity is precipitated with caustic, and the remainder is sorbed on a cation resin, Duolite CS-100. Cesium-137 activity is routinely reduced to 30 Bq/L, and /sup 90/Sr to 0.5 Bq/L. Operation of the plant in 1977 cost $4.64 per thousand gallons (approx. 0.1 cent/L). The routine flow rate is 375 L/min.
Date: January 1, 1982
Creator: Chilton, J.M. & Lasher, L.C.
Partner: UNT Libraries Government Documents Department

Concepts for detritiation of waste liquids

Description: Tritium is formed in thermal nuclear reactors both by neutron activation of elements such as deuterium and lithium and by ternary fission in the fuel. It is a weak beta-emitter with a short half-life, 12.3 years, and its radiological significance in reactor discharges is very low. In heavy-water-cooled and -moderated reactors, such as the SRS reactors, the tritium concentration in the moderator is sufficiently high to cause a potential hazard to operators, so research and development programs have been carried out on processes to remove the tritium. Detritiation of light water has also been the subject of major R D efforts world-wide, because reprocessing operations can generate significant quantities of tritium in liquid waste, and high concentrations of tritium may arise in some aqueous streams in future fusion reactors. This paper presents a review of some of the methods that have been proposed, studied, and developed for removal of tritium from light and heavy water, along with some new concepts for aqueous detritiation directly from liquid oxide (HTO) bearing feed streams.
Date: January 1, 1991
Creator: King, C.M. (Westinghouse Savannah River Co., Aiken, SC (United States)); Van Brunt, V.; Garber, A.R. (South Carolina Univ., Columbia, SC (United States)) & King, R.B. (Georgia Univ., Athens, GA (United States). Dept. of Chemistry)
Partner: UNT Libraries Government Documents Department

DWPF Batch 1, Waste glass investigations

Description: The initial feed to the Defense Waste Processing Facility at the Savannah River Site is currently being prepared and characterized. In the DWPF, this material will be mixed with glass frit and vitrified. The goal of this study is to investigate the effects of variability in the feed mixture on important glass properties. The results will be used to validate the composition -- property models which will be used for process control.
Date: January 1, 1991
Creator: Schumacher, R.F.
Partner: UNT Libraries Government Documents Department

Process for the recovery of curium-244 from nuclear waste

Description: A process has been designed for the recovery of curium from purex waste. Curium and americium are separated from the lanthanides by a TALSPEAK extraction process using differential extraction. Equations were derived for the estimation of the economically optimum conditions for the extraction using laboratory batch extraction data. The preparation of feed for the extraction involves the removal of nitric acid from the Purex waste by vaporization under reduced pressure, the leaching of soluble nitrates from the resulting cake, and the oxalate precipitation of a pure lanthanide-actinide fraction. Final separation of the curium from americium is done by ion-exchange. The steps of the process, except ion-exchange, were tested on a laboratory scale and workable conditions were determined.
Date: October 1, 1980
Creator: Posey, J.C.
Partner: UNT Libraries Government Documents Department

Unclassified information on tritium extraction and purification technology: attachment 1

Description: Several tritium recovery and purification techniques developed at non-production sites are described in the unclassified and declassified literature. Heating of irradiated Li-Al alloy under vacuum to release tritium is described in declassified reports of Argonne National Laboratory. Use of palladium membranes to separate hydrogen isotopes from other gases is described by Argonne, KAPL, and others. Declassified KAPL reports describe tritium sorption on palladium beds and suggest fractional absorption as a means of isotope separation. A thermal diffusion column for tritium enrichment is described in a Canadian report. Mound Laboratory reports describe theoretical and experimental studies of thermal diffusion columns. Oak Ridge reports tabulate ''shape factors'' for thermal diffusion columns. Unclassified journals contain many articles on thermal diffusion theory, experiments, and separation of gas mixtures by thermal diffusion columns; much of these data can be readily extended to the separation of hydrogen-tritium mixtures. Cryogenic distillation for tritium recovery is described in the Mound Laboratory reports. Process equipment such as pumps, valves, Hopcalite beds, and uranium beds are described in reports by ANL, KAPL, and MLM, and in WASH-1269, Tritium Control Technology.
Date: January 23, 1976
Creator: McNorrill, P.L.
Partner: UNT Libraries Government Documents Department

Tritium waste control: April--June 1978. [Catalytic exchange detritiation; liquid waste decontamination; fixation in polymer impregnated concrete; management of high specific activity tritiated wastes]

Description: 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.
Date: July 28, 1978
Partner: UNT Libraries Government Documents Department

Development and selection of a matrix alloy for /sup 85/Kr encapsulation

Description: Pacific Northwest Laboratory has developed and demonstrated a pilot-scale process for stable, long-term storage of radioactive /sup 85/Kr gas from spent nuclear fuel. The process entraps the Kr into a solid metal matrix that can be safely stored at ambient pressure. For this matrix numerous alloys were first screened; those that best satisfied the selection criteria were Cu-Y, Ni-Y, and Ni-La. Of these, Cu-Y alloys containing approximately 20 at.% Y were recommended for use in the pilot-scale system. Reasons for this decision, based on the development work described in Section 5, are summarized here. Thick Cu-Y-Kr deposits (greater than or equal to1 mm) exhibit much better thermal and mechanical stability than do those of Ni-La-Kr and are at least as stable as Ni-Y-Kr deposits. Cu-Y-Kr coatings are very compatible with the sputtering process. They adhere well to the substrate, do not spall significantly during deposition, and can be deposited at higher rates than the Ni-base alloys. This faster deposition helps compensate, in terms of process efficiency, for the lower Kr capacity of Cu-Y-Kr alloys. Another advantage of Cu-Y over Ni-base alloys is the higher vapor pressure of Cu compared to Ni. This reduces the unwanted buildup of Cu on the hot anode surface, whereas deposition of Ni is a problem with Ni-Y, for example. Cu-Y-Kr deposits containing 17 to 20 at. % Y and 6 to 8 at. % Kr compared favorably to Ni/sub 80/La/sub 10/Kr/sub 10/ in terms of long-term Kr retention characteristics. The measurements of Cu-Y-Kr by differential scanning calorimetry also indicated stable retention of Kr because rapid release did not occur below approx.650/sup 0/C. Finally, Cu-Y alloys are satisfactory in terms of materials costs and producibility of the sputtering target. 13 refs., 9 figs., 4 tabs.
Date: July 1, 1986
Creator: Knoll, R.W.; McClanahan, E.D.; Tingey, G.L. & McDonald, E.L.
Partner: UNT Libraries Government Documents Department

Volume reduction system for solid and liquid TRU waste from the nuclear fuel cycle: April--June 1977

Description: The Cyclone Incinerator development this quarter was concentrated on testing a particular liquid feed and burn system. The liquids tested were limited to tributyl phosphate in kerosene and oil-based vacuum pump oil. The tests established that the system worked as designed and built; that tributyl phosphate (TBP), as much as 30% by weight in kerosene, would burn readily; and that the scrubber system collects and neutralizes phosphoric acid which results from the combustion of TBP. Ash immobilization studies included determinations of leach rates for cemented ash. Plutonium-238 was used as a doping agent for the detection of leaching.
Date: November 30, 1977
Creator: Luthy, D.F. & Bond, W.H.
Partner: UNT Libraries Government Documents Department