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Distribution of thorium during full loading of carboxylic acid cation exchange resins with uranium from nitrate solutions at 30 and 40/sup 0/C

Description: The recycle of /sup 233/U for the refabrication of fuel elements for a high-temperature gas-cooled reactor will require an assessment of chemical specifications for impurities in the uranyl nitrate product stream from the fuel reprocessing facility. The behavior of thorium, as an impurity in this simulated process stream, was investigated under equilibrium conditions during full loading of carboxylic acid cation exchange resins with uranium. Mathematical correlations of the data indicate a complex metathetical reaction for thorium distribution between the two phases. Activity coefficient ratios for thorium distribution were calculated by a simple empirical relationship with thorium concentrations in the aqueous phase over the nitrate ion concentration range of approx. 0.2 to 2.0 N. Estimates of thorium behavior and its separation from uranium can be calculated from these values and derived constants for the reaction system. In all tests, thorium was preferentially absorbed on the resin with separation factors, relative to uranium concentrations, greater than unity.
Date: December 1, 1978
Creator: Shaffer, J. H. & Greene, C. W.
Partner: UNT Libraries Government Documents Department

Pyrochemical investigations into recovering plutonium from americium extraction salt residues

Description: Progress into developing a pyrochemical technique for separating and recovering plutonium from spent americium extraction waste salts has concentrated on selective chemical reduction with lanthanum metal and calcium metal and on the solvent extraction of americium with calcium metal. Both techniques are effective for recovering plutonium from the waste salt, although neither appears suitable as a separation technique for recycling a plutonium stream back to mainline purification processes. 17 refs., 13 figs., 2 tabs.
Date: May 1, 1987
Creator: Fife, K.W. & West, M.H.
Partner: UNT Libraries Government Documents Department

Ion exchange flowsheet for recovery of cesium from purex sludge supernatant at B Plant

Description: Purex Sludge Supernatant (PSS) contains significant amounts of /sup 137/Cs left after removal of strontium from fission product bearing Purex wastes. To remove cesium from PSS, an Ion Exchange Recovery system has been set up in Cells 17-21 at B Plant. The cesium that is recovered is stored within B Plant for eventual purification through the Cesium Purification process in Cell 38 and eventual encapsulation and storage in a powdered form at the Waste Encapsulation Storage Facility. Cesium depleted waste streams from the Ion Exchange processes are transferred to underground storage.
Date: April 29, 1977
Creator: Carlstrom, R.F.
Partner: UNT Libraries Government Documents Department

Krypton-85 storage in solid matrices. [By ion implantation during sputtering, or by dissolution in glass]

Description: Storage of Kr-85 will be required in support of nuclear power reactors beginning in 1983. Both approaches described here appear to meet the requirements for such a storage medium. Entrapment of the Kr during sputtering has several rather obvious safety advantages. The operation of the process at low rho and at or below room temperature should reduce markedly the potential for significant Kr-85 release to the environment during processing of the waste stream. It also appears that adaptation of this process for handling radioactive materials would also be simpler than the large high pressure, high temperature apparatus required for loading the glass sample. Furthermore, a significantly higher Kr loading is possible in the sputtered metals thus reducing the volume required for storage by as much as a factor of 50 to 100. On the other hand, the low density loaded glass process takes advantage of a very inexpensive starting material and existing commercial technology for high temperature, high pressure processes. The volume of the Kr-loaded glass matrix could be reduced by going to still higher pressures.
Date: January 1, 1979
Creator: Tingey, G. L.; McClanahan, E. D.; Bayne, M. A.; Gray, W. J. & Hinman, C. A.
Partner: UNT Libraries Government Documents Department

Reference thorium fuel cycle

Description: In the reference fuel cycle for the TFCT program, fissile U will be denatured by mixing with /sup 238/U; the plants will be located in secure areas, with Pu being recycled within these secure areas; Th will be recycled with recovered U and Pu; the head end will handle a variety of core and blanket fuel assembly designs for LWRs and HWRs; the fuel may be a homogeneous mixture either of U and Th oxide pellets or sol-gel microspheres; the cladding will be Zircaloy; and MgO may be added to the fuel to improve Th dissolution. Th is being considered as the fertile component of fuel in order to increase proliferation resistance. Spent U recovered from Th-based fuels must be re-enriched before recycle to prevent very rapid buildup of /sup 238/U. Stainless steel will be considered as a backup to Zircaloy cladding in case Zr is incompatible with commercial aqueous dissolution. Storage of recovered irradiated Th will be considered as a backup to its use in the recycle of recovered Pu and U. Estimates are made of the time for introducing the Th fuel cycle into the LWR power industry. Since U fuel exposures in LWRs are likely to increase from 30,000 to 50,000 MWD/MT, the Th reprocessing plant should also be designed for Th fuel with 50,000 MWD/MT exposure. (DLC)
Date: August 1, 1978
Creator: Driggers, F.E.
Partner: UNT Libraries Government Documents Department

Shuffler instruments for the nondestructive assay of fissile materials

Description: A shuffler is a nondestructive assay instrument used to determine the fissile content of materials. It places an isotopic source of neutrons near the material to induce fissions, withdraws the source, and counts the delayed neutrons. The source is shuffled until a sufficient number of delayed neutrons have been counted. The shuffler technique is generally applied to difficult assay cases. The amount of material present may be very small (a few milligrams), and thus it does not spontaneously emit neutrons of consequence; the amount of material is also below an active well counter's level of sensitivity. On the other hand, the fissile amount may be fairly large, but the rate of spontaneously emitted neutrons may still be low (so a passive neutron count will not work) or the highest assay precision may be desired (favoring a shuffler over an active well counter) even if the material is inhomogeneous (making it difficult to interrogate with thermal neutrons). In all these cases, gamma-ray backgrounds, self- shielding, or matrix effects can make gamma-ray assays impractical. Materials ranging from highly radioactive spent-fuel assemblies to low-level waste drums have been assayed with shufflers, as have leached hulls, various process materials, scrap, and waste. This report presents a theoretical background for shufflers and describes techniques for practical applications. Procedures for assaying mixtures of fissile isotopes, inhomogeneous materials, and flowing liquids are discussed. It is shown how the precision and limits of detection of a shuffler can be calculated for a given neutron background rate. A section on data analysis gives a stepwise procedure for converting the measured counts into an assay value, including random, systematic, and total uncertainties. 31 refs.
Date: May 1, 1991
Creator: Rinard, P.M.
Partner: UNT Libraries Government Documents Department

Evaluation of target power supplies for krypton storage in sputter-deposited metals

Description: Implantation of /sup 85/Kr in a growing sputtered metal deposit has been studied for the containment of /sup 85/Kr recovered from the reprocessing of spent nuclear fuel. PNL, as part of DOE's research program for /sup 85/Kr storage, has developed krypton trapping storage devices (KTSDs) in a range of sizes for ''cold'' and radioactive testing. The KTSD is a stainless steel canister that contains a sputtering target for depositing an amorphous rare-earth transition metal on the inner wall and simultaneously implanting low-energy krypton ions in the growing deposit. This report covers the design requirements for the target power supply and the description, testing and evaluation of three basic designs. The designs chosen for evaluation were: (1) a standard commercial power supply with an external PNL-designed current interrupter, (2) a commercially manufactured power supply with an integral series-type interrupter, and (3) a commercially manufactured power supply with an integral shunt-type interrupter. The units were compared on the basis of performance, reliability, and life-cycle cost. 8 refs., 9 figs., 2 tabs.
Date: April 1, 1986
Creator: Greenwell, E.N.; McClanahan, E.D. & Moss, R.W.
Partner: UNT Libraries Government Documents Department

Characterization of airborne plutonium-bearing particles from a nuclear reprocessing plant

Description: The elemental compositions, sizes, structures, and /sup 239/Pu contents were determined for 558 plutonium-bearing particles isolated from airborne particles collected at various locations in the exhaust from a nuclear fuel reprocessing facility. These data were compared with data from natural aerosol particles. Most of the collected particles were composed of aggregates of crustal materials. 3.6% of the particles were organic and 1.7% were metallic, viz., iron, chromium, and nickel. High enrichment factors for titanium, manganese, chromium, nickel, zinc, and copper were evidence of the anthropogenic nature of some of the particles. Plutonium contents of most particles were very low (less than one femtocurie of /sup 239/Pu). Plutonium concentrations were determined by the fission track counting method. Only one particle contained sufficient plutonium for detection by electron microprobe analysis. This was a 1-..mu..m diameter particle containing 73% PuO/sub 2/ by weight (estimated to be 170 fCi of /sup 239/Pu) in combination with Fe/sub 2/O/sub 3/ and mica. The plutonium-bearing particles were generally larger than natural aerosols. The geometric mean diameter of those collected from the mechanical line exhaust point where plutonium is converted to the metal was larger than that of particles collected from the wet cabinet exhaust (12.3 ..mu..m vs 4.6 ..mu..m). Particles from the mechanical line also contained more plutonium per particle than those from the wet cabinets. The amount of plutonium per particle decreased with the distance of each sampling point from the mechanical line which is considered the major source of plutonium contamination in the reprocessing facility.
Date: January 1, 1978
Creator: Sanders, S.M.
Partner: UNT Libraries Government Documents Department

Voloxidation and dissolution of irradiated (Th,U)O/sub 2/

Description: Exploratory hot-cell tests were made to determine the amounts of fission products released by high-temperature oxidation (voloxidation) of irradiated fuel, initially (Th/sub 0/ /sub 96/U/sub 0/ /sub 04/)O/sub 2/, and whether the treatment would affect the subsequent dissolution of the oxide. Fifteen-year-decayed fuel rods were sheared to dislodge the (Th,U)O/sub 2/ from its cladding, sieved to determine the size range, and an aliquot was oxidized in flowing air at 600/sup 0/C for approx. 2 h while tumbling at 12 rpm. The voloxidation released only 0.2% of the /sup 85/Kr and 6% of the /sup 3/H contained in the (Th,U)O/sub 2/. The oxidized and unoxidized portions were leached three times in a 95/sup 0/C Thorex reagent to dissolve the fuel. The voloxidation apparently had no effect on the dissolution of the mixed oxide. The presence of cladding slowed the dissolution by limiting the amount of fuel exposed to the acid.
Date: April 1, 1979
Creator: Goode, J.H. & Stacy, R.G.
Partner: UNT Libraries Government Documents Department

Actinide partitioning-transmutation program final report. I. Overall assessment

Description: This report is concerned with an overall assessment of the feasibility of and incentives for partitioning (recovering) long-lived nuclides from fuel reprocessing and fuel refabrication plant radioactive wastes and transmuting them to shorter-lived or stable nuclides by neutron irradiation. The principal class of nuclides considered is the actinides, although a brief analysis is given of the partitioning and transmutation (P-T) of /sup 99/Tc and /sup 129/I. The results obtained in this program permit us to make a comparison of the impacts of waste management with and without actinide recovery and transmutation. Three major conclusions concerning technical feasibility can be drawn from the assessment: (1) actinide P-T is feasible, subject to the acceptability of fuels containing recycle actinides; (2) technetium P-T is feasible if satisfactory partitioning processes can be developed and satisfactory fuels identified (no studies have been made in this area); and (3) iodine P-T is marginally feasible at best because of the low transmutation rates, the high volatility, and the corrosiveness of iodine and iodine compounds. It was concluded on the basis of a very conservative repository risk analysis that there are no safety or cost incentives for actinide P-T. In fact, if nonradiological risks are included, the short-term risks of P-T exceed the long-term benefits integrated over a period of 1 million years. Incentives for technetium and iodine P-T exist only if extremely conservative long-term risk analyses are used. Further RD and D in support of P-T is not warranted.
Date: June 1, 1980
Creator: Croff, A.G.; Blomeke, J.O. & Finney, B.C.
Partner: UNT Libraries Government Documents Department

Consolidated fuel reprocessing program. Developments for the future in reprocessing

Description: The future reprocessing developments focus on three major areas: (1) the retention of gaseous fission products to reduce off-site doses to very low values; (2) the initial steps of breakdown, shearing, and dissolution of breeder fuels; and (3) advanced facility and equipment concepts, which are expected to lead to a reliable, cost-effective, totally remotely operated and maintained plant. Work in the first area - removal of fission gases (the most important of which is /sup 85/Kr) - is largely completed through tracer and bench-scale engineering equipment. Efforts are now mainly devoted to breeder fuels and advanced remote concepts. A facility, the Integrated Equipment Test Facility, which will be used to carry out much of this work, is nearing completion in Oak Ridge. In it a large, simulated, remote reprocessing cell will house a disassembly-shear machine for either breeder or LWR fuels, a rotary continuous dissolver, a solvent extraction cycle utilizing a new generation of centrifugal contactors, and related equipment.
Date: January 1, 1982
Creator: Burch, W.D.
Partner: UNT Libraries Government Documents Department

Voloxidation studies with UO/sub 2/ reactor fuels

Description: Voloxidation was studied experimentally in small-scale laboratory tests with irradiated UO/sub 2/ reactor fuels. The present study developed new data on the effects on the voloxidation reaction of hull length, oxygen concentration, temperature, agitation, fuel density, and fuel type and burnup. The reaction was studied by measuring weight gains, yields of U/sub 3/O/sub 8/, product densities and particle-size distributions, reaction rates, tritium release, and behavior of other off-gases (/sup 14/C, /sup 85/Kr, and /sup 129/I).
Date: January 1, 1980
Creator: Stone, J A
Partner: UNT Libraries Government Documents Department

Hot cell studies of tritium removal from and dissolution of an irradiated thoria fuel

Description: Tritium removal and dissolution studies on an irradiated thoria-based fuel were conducted in the High Level Caves at the Savannah River Laboratory (SRL). The objectives of these studies were to define the effects of key process-related parameters on tritium evolution and subsequent dissolution. The test program at SRL determined the effects on tritium removal of particle size, heating temperature, oxidation, and agitation. ThO/sub 2//UO/sub 2/ (95%/5%) fuel from the Elk River Reactor, irradiated to about 12,000 MWD/MTHM and cooled for about 12 years, was used in the tests. The thoria/urania fuel pellets were separated from the stainless steel cladding and were divided into size fractions to determine the particle-size distribution resulting from the decladding process. In the tritium removal tests, the effect of heating several different particle-size fractions was studied at temperatures ranging from 600 to 1000/sup 0/C in the presence of air for 10 to 30 hours. Each of the roasted samples (about 100 grams) was subsequently dissolved in refluxing Thorex reagent (13M HNO/sub 3/, 0.05M HF, and 0.1M Al(NO/sub 3/)/sub 3/) to determine the residual tritium. The amount of insoluble residue remaining after the dissolution was determined and characterized.
Date: January 1, 1980
Creator: Pickett, J.B.
Partner: UNT Libraries Government Documents Department

Review of thorium fuel reprocessing experience

Description: The review reveals that experience in the reprocessing of irradiated thorium materials is limited. Plants that have processed thorium-based fuels were not optimized for the operations. Previous demonstrations of several viable flowsheets provide a sound technological base for the development of optimum reprocessing methods and facilities. In addition to the resource benefit by using thorium, recent nonproliferation thrusts have rejuvenated an interest in thorium reprocessing. Extensive radiation is generated as the result of /sup 232/U-contamination produced in the /sup 233/U, resulting in the remote operation and fabrication operations and increased fuel cycle costs. Development of the denatured thorium flowsheet, which is currently of interest because of nonproliferation concerns, represents a difficult technological challenge.
Date: January 1, 1978
Creator: Brooksbank, R.E.; McDuffee, W.T. & Rainey, R.H.
Partner: UNT Libraries Government Documents Department

Light water reactor fuel reprocessing: dissolution studies of voloxidized fuel

Description: Voloxidation is a proposed head-end process to remove tritium from irradiated LWR fuel by roasting the fuel in the presence of oxygen. The process oxidizes UO/sub 2/ to a fine U/sub 3/O/sub 8/ powder with high surface area for dissolution. Small-scale tests with irradiated Robinson, Oconee, and Saxton reactor fuels have been made to determine the dissolution behavior of both voloxidized and nonvoloxidized (UO/sub 2/) fuel. No significant technical problems were encountered in batch-dissolving of the U/sub 3/O/sub 8/ or UO/sub 2/ fuel. Dissolution rates were well-controlled in all tests. Significant observations from U/sub 3/O/sub 8/ dissolution, when compared to UO/sub 2/ dissolution, included: (1) reduced tritium and ruthenium (/sup 106/Ru) concentration in dissolver solutions, (2) increased weight of insoluble noble metal fission product residue (approximately 2.2X greater), and (3) increased fraction of the total plutonium which remains insoluble and is collected with the fission product residue. The insoluble plutonium was leached easily from the residue by 10M HNO/sub 3/ to ensure quantitative plutonium recovery. The weight of the fission product residue collected from both U/sub 3/O/sub 8/ and UO/sub 2/ fuels increased linearly with fuel burnup. A major fraction (> 88%) of the /sup 85/Kr was evolved from U/sub 3/O/sub 8/ fuel during dissolution rather than voloxidation. The /sup 85/Kr evolution rate was an appropriate monitor of fuel dissolution rate. Virtually all of the /sup 129/I was evolved by air sparging of the dissolver solution during dissolution.
Date: January 1, 1977
Creator: Johnson, D.R. & Stone, J.A.
Partner: UNT Libraries Government Documents Department

Computer-controlled on-line gamma analysis for krypton-85

Description: /sup 85/Kr will be evolved from spent nuclear fuel during both the voloxidation and dissolution processes, so a reliable method for on-line analysis of /sup 85/Kr in the off-gas system is needed. Tritium, /sup 14/C, and /sup 129/I were trapped, and the activity of /sup 85/Kr was then measured using a Li-drifted Ge detector. Equipment used to carry out this analysis is described; the PET computer is used. The /sup 85/Kr evolution rate was correlated with the fuel dissolution rate; the close correlation permits one to monitor the fuel dissolution process. 11 figures. (DLC)
Date: March 1, 1980
Creator: Canuette, R.P.
Partner: UNT Libraries Government Documents Department

Study of the formation, prevention, and recovery of plutonium from plutonium esters in the Purex process

Description: The Savannah River Plant uses the basic Purex process to separate /sup 239/Pu from /sup 238/U and fission products. Dark-brown, dense solids containing up to 30% Pu have previously occurred in rotameters in the plutonium finishing operations. The kinetics of formation of this mixture of DBP- and MBP-Pu esters suggest two methods to prevent the formation of the solids. A selective dissolution method using NaOH metathesis has been developed to separate the phosphate ester from the plutonium before dissolution of the residual plutonium hydroxide in a HNO/sub 3/-HF medium.
Date: January 1, 1981
Creator: Gray, L. W. & Burney, G. A.
Partner: UNT Libraries Government Documents Department

Calculated critical parameters in simple geometries for oxide and nitrate water mixtures of U-233, U-235 and Pu-239 with thorium. Final report

Description: Calculations have been performed on water mixtures of oxides and nitrates of /sup 233/U, /sup 235/U, and /sup 239/Pu with chemically similar thorium compounds to determine critical dimensions for simple geometries (sphere, cylinder, and slab). Uranium enrichments calculated were 100%, 20%, 10%, and 5%; plutonium calculations assumed 100% /sup 239/Pu. Thorium to uranium or plutonium weight ratios (Th: U or Pu) calculated were 0, 1, 4, and 8. Both bare and full water reflection conditions were calculated. The results of the calculations are plotted showing a critical dimension versus the uranium or plutonium concentration. Plots of K-infinity and material buckling for each material type are also shown.
Date: November 1, 1979
Creator: Converse, W.E. & Bierman, S.R.
Partner: UNT Libraries Government Documents Department

Application of on-line plutonium isotopic concentration monitors at a nuclear fuel reprocessing plant

Description: Gamma-ray spectroscopy is used to assay plutonium solutions flowing through thin sample cells. The direct measurement of gamma-ray intensities are used to determine the plutonium isotopic content and total concentration. The development of spectral analysis algorithms, the use of appropriate instrument internal standards, techniques to minimize spectral acquisition times, and methods to deduce the /sup 242/Pu content are discussed. Results of experiments on plutonium solutions in the 200 to 500 grams/liter concentration range of typical light water reactor isotopic content are given. These solutions typify the purified plutonium product streams in a reprocessing plant. The isotopic and total plutonium assay results as determined by this technique are within 0.5% of the reference values, as determined by mass spectrometry and controlled potential coulometry, respectively. Similar techniquescan also be applied to low concentration plutonium solutions. Plutonium isotopic ratios can be determined by this technique and the alpha specific activities computed. For low concentration (10/sup -0/ to 10/sup -4/ grams/liter) streams, cerium activated Vycor scintillation detectors can then be used for plutonium monitoring. This is compared with other proposed methods of on-line plutonium analyses. Modes of operation of these monitors for real-time inventory and diversion detection are discussed.
Date: February 1, 1979
Creator: Hofstetter, K.J. & Huff, G.A.
Partner: UNT Libraries Government Documents Department

Laboratory studies on the evolution of iodine-129 during Purex-uranium metal dissolution

Description: The path of iodine from the Purex dissolver was determined during fuel dissolution using /sup 125/I tracer. Laboratory-scale equipment qualification studies were completed using sections of nonirradiated uranium N-reactor fuel elements. A proof-of-principle dissolution study was completed at the end of FY 1979 in the PNL hot cells using wafers of irradiated N-reactor fuel. The findings include the following: the laboratory-scale dissolver/downdraft condenser was designed at a factor of 5 x 10/sup -5/ of the Purex flowsheet; with no refluxing, 5.6 moles of HNO/sub 3/ were required per mole of dissolved uranium. With NO/sub x/ recovery in the reflux stream, an average of 3.6 moles of HNO/sub 3/ was required. These results formed the basis for adequate modeling of the laboratory Purex downdraft dissolver; approximately 8% of the iodine was found in the final dissolver solution when the /sup 125/I tracer was added to the initial dissolver solution prior to the first cut, 6-h dissolution; when the /sup 125/I was added continuously during the 6-h dissolution without any refluxing of the condenser acid back to the dissolver, approximately 16% of the iodine was found in the dissolver solution; when irradiated N-reactor fuel was dissolved while /sup 125/I tracer was continuously added to the dissolver during the 6-h test, 11% of the /sup 125/I tracer was found in the dissolver solution. After 2 h of refluxing with air sparging, 6% of the /sup 125/I tracer was found in the dissolver solution; and analysis of the fission product /sup 129/I in the fuel duplicated the tracer study results with 8% and 7% of the iodine remaining in the dissolver solution after 6 and 8 h, respectively.
Date: March 1, 1980
Creator: Bray, L.A.
Partner: UNT Libraries Government Documents Department

Recovery of plutonium from molten salt extraction residues

Description: Savannah River Laboratory (SRL), Savannah River Plant (SRP), and Rocky Flats Plant (RFP) are jointly developing a process to recover plutonium from molten salt extraction residues. These NaCl, KCl, MgCl/sub 2/ residues, which are generated in the pyrochemical extraction of /sup 241/Am from aged plutonium metal, contain up to 25 wt % dissolved PUCl/sub 3/ and up to 2 wt % AmCl/sub 3/. The objective is to develop a process to convert these residues to plutonium metal product and discardable waste. The first step of the conceptual process is to convert the actinides to a heterogenous scrub alloy with aluminum and magnesium. This step, performed at RFP, effectively separates the actinides from the bulk of the chloride. This scrub alloy will then be dissolved in a HNO/sub 3/-HF solution at SRP. Residual chloride will be removed by precipitation with Hg/sub 2/(NO/sub 3/)/sub 2/ followed by centrifugation. Plutonium and americium will be separated using the Purex solvent extraction process. The /sup 241/Am will be diverted to the solvent extraction waste stream where it can either be discarded to the waste farm or recovered. The plutonium will be finished via PuF/sub 3/ precipitation, oxidation to a mixture of PUF/sub 4/ and PuO/sub 2/, followed by reduction to plutonium metal with calcium.
Date: January 1, 1983
Creator: Gray, L.W. & Holcomb, H.P.
Partner: UNT Libraries Government Documents Department

Coordination chemistry of several radius-sensitive complexones and applications to lanthanide-actinide separations

Description: The relationships between the lanthanide complex formation equilibria and the lanthanide-actinide separation application of three radius sensitive ligands have been studied. The consecutive stepwise formation constants of the 1:1, 2:1, and 3:1 chelate species formed by the interaction of DHDMB and the tripositive lanthanides and yttrium were determined potentiometrically at 0.1 M ionic strength and 25/sup 0/C. Results indicate that three different coordination modes, one tridentate and two bidentate are in evidence. Tracer level /sup 241/Am - /sup 155/Eu cation-exchange experiments utilizing DHDMB eluents indicate that this dihydroxycarboxylate does not form a sufficiently strong americium complex to elute that actinide ahead of europium. The overall stability of the americium 3:1 complex appears intermediate between samarium and europium. Cation-exchange elutions of /sup 241/Am, /sup 155/Eu, and /sup 160/Tb mixtures with EEDTA solutions prove that the EEDTA ligand is capable of eluting americium ahead of all of the tripositive lanthanide cations. The minimum separation occurs with terbium, where the Am-Tb separation factor is 1.71. 1,5-diaminopentane-N,N,N',N'-tetraacetic acid (PMDTA) was synthesized using cation exchange. A mathematical method was developed for the formation constants of the protonated and unprotonated lanthanide-PMDTA complexes from potentiometry. Cation-exchange elutions of tracer quantities of Am, Eu, and Tb revealed that terbium is eluted ahead of both americium and europium.
Date: October 1, 1981
Creator: Potter, M.W.
Partner: UNT Libraries Government Documents Department

Comparison of silver sorbents for application to radioiodine control at the PUREX process facility modification. [Iodine 129]

Description: In continued support of the design of the gaseous radioiodine control system for the PUREX Process Facility Modification (PFM), the Pacific Northwest Laboratory (PNL) conducted laboratory-scale measurements of the performance of four state-of-the-art sorbents for radioiodine in the dissolver offgas (DOG) of a nuclear reprocessing plant. The PFM is a new head-end treatment plant being designed by Westinghouse Hanford Company (WHC) for the PUREX Plant at the Hanford Site. The experiments performed measured the iodine effluent concentration from Norton silver mordenite (NAgZ), Linde silver mordenite (LAgZ), Linde silver faujasite (AgX), and silver nitrate-impregnated silicic acid (AgNO/sub 3/Si) during simulated normal operating conditions in the PFM after three shutdown/startup cycles, and during standby. At normal operating conditions the input gas is expected to have a dew point of 35/degree/C to 40/degree/C and contain 0.1 ..mu..mol I/L, 1 vol% NO, and 1 vol% NO /sub 2/. The sorbent bed would be at 150/degree/C. A shutdown/startup cycle consisted of eliminating iodine and NO/sub x/ from the input gas, cooling the bed to room temperature, stopping gas flow, and restarting the system. During standby conditions the input gas contained no iodine or NO/sub x/, the dew point was at 30/degree/C to 35/degree/C, and the bed temperature remained at 150/degree/C. This experimental study showed that 20 cm beds of NAgZ, LAgZ, and 18 wt% silver AgX could load up to 0.25 mmol I/g sorbent and routinely reduce the iodine concentration in a simulated PFM DOG from 0.1 ..mu..mol I/L to less than the target level of 10/sup /minus/5/ ..mu..mol I/L. In contrast, the AgNO/sub 3/Si unexpectedly failed to achieve this required level of performance, reducing the concentration on a routine basis only to 10/sup /minus/4/ to 10/sup /minus/2/ ..mu..mol I/L. 5 refs., 14 figs., 6 tabs.
Date: September 1, 1988
Creator: Scheele, R.D.; Burger, L.L. & Halko, B.T.
Partner: UNT Libraries Government Documents Department

Development of Ag/sup 0/Z for bulk /sup 129/I removal from nuclear fuel reprocessing plants and PbX for /sup 129/I storage

Description: Tests were conducted to develop Ag-exchanged mordenite (AgZ) for removal of gaseous /sup 129/I from nuclear fuel reprocessing plants. The effects of bed depth and hydrogen pretreatment on the elemental (I/sub 2/) iodine loading of AgZ were examined. The tests indicated that reduced AgZ (Ag/sup 0/Z) had about twice the capacity for iodine as AgZ, and at least 15-cm bed depths should be used for loading tests. The effects of H/sub 2/O(g), NO, NO/sub 2/, and bed temperature on the iodine loading of Ag/sup 0/Z were determined. The highest loadings were obtained with NO in the gas stream. Water vapor and bed temperature appeared to have no effect on the iodine loadings. Tests were conducted to develop a dry method for in situ regeneration of iodine-loaded Ag/sup 0/Z. A test bed of Ag/sup 0/Z was recycled 13 times by loading it with I/sub 2/ and stripping the I/sub 2/ (as HI) with H/sub 2/. A 20 percent loss in iodine capacity was observed by the fourteenth loading. The iodine loadings of lead-exchanged zeolites, which were used to chemisorb HI during the recycle tests, were measured. The iodine vapor pressures at 20/sup 0/C over the substrates were predicted to be 10/sup -6/, 10/sup -8/, and less than 10/sup -16/ atm for Pb-exchanged mordenite, Pb-exchanged faujasite and reduced Pb-exchanged faujasite, respectively. A process-flow diagram was formulated for iodine recovery. Two parallel beds of Ag/sup 0/Z are used to permit continuous iodine recovery. While one is being regenerated the other recovers iodine. The iodine is chemisorbed as AgI, stripped as HI, and chemisorbed on the lead bed as a form of PbI/sub 2/. The Ag beds were sized for a 30-day operation in a fuel reprocessing plant before regeneration. About 5 days would be needed for regeneration. About 2 m/sup 3/ of iodine-loaded ...
Date: January 1, 1978
Creator: Thomas, T.R.; Staples, B.A. & Murphy, L.P.
Partner: UNT Libraries Government Documents Department