47 Matching Results

Search Results

Advanced search parameters have been applied.

Production of Depleted UO<sub>2</sub>Kernels for the Advanced Gas-Cooled Reactor Program for Use in TRISO Coating Development

Description: The main objective of the Depleted UO{sub 2} Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO{sub 2} kernels with diameters of 500 {+-} 20 {micro}m and 3.5 kg of UO{sub 2} kernels with diameters of 350 {+-} 10 {micro}m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO{sub 2} kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO{sub 3} {center_dot} 2H{sub 2}O microspheres to form dense UO{sub 2} kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 {+-} 10-{micro}m-diameter kernels, and to obtain very high yields.
Date: December 2, 2004
Creator: Collins, J. L.
Partner: UNT Libraries Government Documents Department

Control of Urania Crystallite Size by HMTA-Urea Reactions in the Internal Gelation Process for Preparing (U, Pu)O<sub>2</sub>Fuel Kernels

Description: In the development of (U,Pu)O{sub 2} kernels by the internal gelation process for the Direct Press Spheroidized process at Oak Ridge National Laboratory, a novel crystal growth step was discovered that made it possible to prepare calcined porous kernels that could be used as direct-press feed for Fast Breeder Reactor pellet fabrication. High-quality pellets were prepared that were near theoretical density and that (upon examination) revealed no evidence of sphere remnants. The controlled crystal growth step involved using hexamethylenetetramine (HMTA)-urea stock solutions that were boiled for 60 min or less. Before this discovery, all the other crystal growth steps (when utilized) could reduce the tap density to only {approx}1.3 g/cm{sup 3}, which was not sufficiently low for use in ideal pellet pressing. The use of the boiled HMTA-urea solution allowed the tap density to be lowered to 0.93 g/cm{sup 3}, with the ideal density being about 1.0 g/cm{sup 3}. This report describes the development of this technology and its scaleup.
Date: April 26, 2005
Creator: Collins, J.L.
Partner: UNT Libraries Government Documents Department

Experimental Methodology for Determining Optimum Process Parameters for Production of Hydrous Metal Oxides by Internal Gelation

Description: The objective of this report is to describe a simple but very useful experimental methodology that was used to determine optimum process parameters for preparing several hydrous metal-oxide gel spheres by the internal gelation process. The method is inexpensive and very effective in collection of key gel-forming data that are needed to prepare the hydrous metal-oxide microspheres of the best quality for a number of elements.
Date: October 28, 2005
Creator: Collins, J.L.
Partner: UNT Libraries Government Documents Department

Development of Spheroidal Inorganic Sorbents for Treatment of Acidic Salt-Bearing Liquid Waste

Description: A spheroidal composite inorganic sorbent was developed for U.S. Department of Energy-Efficient Separations and Processing Crosscutting Program (USDOE-ESP) for potential use in removing radioactive cesium isotopes from acidic high-salt waste streams such as those at Idaho National Engineering and Environmental Laboratory (INEEL). The sorbent, zirconium monohydrogen phosphate (ZrHP) embedded with fine powder of ammonium molybdophosphate (AMP), was prepared using a unique internal gelation process which can be used to make porous reproducible microspheres that are structurally strong, have a low tendency for surface erosion, and improve the flow dynamics for column operations. Both ZrHP and AMP are excellent sorbent materials and, being inorganic, are stable in high radiation fields. AMP is a very effective sorbent for removing cesium from salt-bearing waste streams for a wide range of acidity. In the pH range of 2 to 10, ZrHP is also a very effective sorbent for removing Cs, Sr, Th, U(VI), Pu(IV), Am(III), Hg, and Pb from streams of lower ionic concentrations. Crucial to developing the spheroidal AMP-ZrHP sorbent was to determine the ideal weight percentage of AMP that could be embedded in the ZrHP microspheres in order to maintain the structural integrity of the microspheres and also achieve a good cesium separation. A total of 12 preparations were made. The dry weight percentage of AMP ranged from 30 to 60. Overall, the best composite microspheres prepared contained 50% AMP (by dry weight measurement). Another composite microsphere, which was composed of titanium monohydrogen phosphate (TiHP) embedded with 18 wt % (air-dried weight) potassium cobalt hexacyanoferrate (KCoCF) and developed for a different separations application, was also batch tested for comparison. It proved to be as effective in removing,the cesium as the air-dried AMP (50 wt %)-ZrHP. Granular KCoCF was also prepared and was very effective. Large samples of each of these materials were ...
Date: September 7, 2001
Creator: Collins, J.L.
Partner: UNT Libraries Government Documents Department

Economic Evaluation for the Production of Sorbents and Catalysts Derived from Hydrous Titanium Oxide Microspheres Prepared by the HMTA Internal Gelation Process

Description: Hydrous metal oxides of Zr, Ti, Hf, Fe, Al, etc. are inorganic ion exchangers that have high selectivities and efficiencies for separating and removing fission products, actinides, and other undesirable elements from aqueous waste streams. In most cases, these ion exchangers are commercially available only as fine powders or as unstable granular particles that are not readily adaptable to continuous processing techniques such as column chromatography. Hydrous metal oxides can be prepared as microspheres by the internal gelation process. This process is unique in that it provides a means of making a usable engineered form of inorganic ion exchanger that can be used in large-scale column separations. With such material, the flow dynamics in column operations would be greatly enhanced. In addition, the microspheres are in a stable form that has little or no tendency to degrade under dynamic conditions. Another advantage of the process is that the gelation time and size of the microspheres can be controlled. Also, microspheres can be reproducibly prepared on either a small or a large scale-which is not always true for batch preparation of the powdered or granular forms. The use of these materials can be expanded in a number of ways. The process allows for the microspheres to be homogeneously embedded with fine particles of other selective ion exchangers, and for the microspheres (undried) to be chemically converted to microspheres of other ion-exchanger materials such as phosphates, silicophosphates, hexacyanoferrates, tungstates, and molybdates. This report presents an economic evaluation of the preparation of hydrous titanium oxide (HTiO) microspheres by an internal gelation process for use in making ion exchangers, catalysts, and getters. It also examines the estimated costs for a company to produce the material but does not discuss the price to be charged since that value would take into account company policy-matters that ...
Date: January 11, 2001
Creator: Collins, J.L.
Partner: UNT Libraries Government Documents Department

Release and transport of fission product cesium in the TMI-2 accident

Description: Approximately 50% of the fission product cesium was released from the overheated UO/sub 2/ fuel in the TMI-2 accident. Steam that boiled away from a water pool in the bottom of the reactor vessel transported the released fission products throughout the reactor coolant system (RCS). Some fission products passed directly through a leaking valve with steam and water into the containment structure, but most deposited on dry surfaces inside of the RCS before being dissolved or resuspended when the RCS was refilled with water. A cesium transport model was developed that extended measured cesium in the RCS back to the first day of the accident. The model revealed that approx.62% of the released /sup 137/Cs deposited on dry surfaces inside of the RCS before being slowly leached and transported out of the RCS in leaked or letdown water. The leach rates from the model agreed reasonably well with those measured in the laboratory. The chemical behavior of cesium in the TMI-2 accident agreed with that observed in fission product release tests at Oak Ridge National Laboratory (ORNL).
Date: January 1, 1986
Creator: Lorenz, R.A. & Collins, J.L.
Partner: UNT Libraries Government Documents Department

Fission product source terms for the LWR loss-of-coolant accident

Description: Models for cesium and iodine release from light-water reactor (LWR) fuel rods failed in steam were formulated based on experimental fission product release data from several types of failed LWR fuel rods. The models were applied to a pressurized water reactor (PWR) undergoing a hypothetical loss-of-coolant accident (LOCA) temperature transient. Calculated total iodine and cesium releases from the fuel rods were 0.053 and 0.025% of the total reactor inventories of these elements, respectively, with most of the release occurring at the time of rupture. These values are approximately two orders of magnitude less than releases used in WASH-1400, the Reactor Safety Study.
Date: July 1, 1980
Creator: Lorenz, R.A.; Collins, J.L. & Malinauskas, A.P.
Partner: UNT Libraries Government Documents Department

Fission product release from defected LWR fuel rods

Description: Experiments conducted at Oak Ridge National Laboratory both with fission product simulants and with irradiated commercial fuel have been utilized to develop a semi-empirical model of fission product release from defected Light Water Reactor (LWR) fuel rods. At fuel temperatures less than 1200/sup 0/C, releases occur from fission products previously accumulated in the pellet-to-cladding gap region. In this temperature range, the release of species of moderate volatility is postulated to result from two processes. The first of these, which occurs during the period of fuel clad rupture, is due to the transport of the fill and fission product gases as they are vented through the cladding defect. The second mechanism for release, which is time-dependent, involves the diffusional transport of the semi-volatile species to the point of clad rupture through the interconnected voids (the pellet-to-cladding gap and cracks in fuel pellets) within the fuel rod.
Date: January 1, 1979
Creator: Malinauskas, A.P.; Lorenz, R.A. & Collins, J.L.
Partner: UNT Libraries Government Documents Department

Behavior of Cs, I, and Te in the fission product release program at ORNL

Description: Experiments have been conducted at ORNL with highly irradiated light-water reactor (PWR and BWR) fuel rod segments to investigate fission product release in steam in the temperature range 500 to 2000/sup 0/C. Objectives were to quantify and characterize the releases under conditions postulated for LOCA) and severe accident conditions. In all, 26 experiments have been conducted - 24 with high burnup and 2 with low burnup fuels. To aid in the interpretation of fission product release, 12 implant and 18 control experiments were also conducted; the behavior of HI, I/sub 2/, Cs/sub 2/O, CsOH, Te, and TeO/sub 2/ (individually and in different combinations) was studied. This paper discusses only the observed behavior of cesium, iodine, and tellurium. Cs and I were released primarily as CsOH and CsI, and Te release was controlled by steam oxidation of Zircaloy cladding.
Date: January 1, 1984
Creator: Collins, J.L.; Osborne, M.F. & Lorenz, R.A.
Partner: UNT Libraries Government Documents Department

Behavior of fission product tellurium under severe accident conditions

Description: Fission product release tests at Oak Ridge National Laboratory (ORNL) have provided new experimental data that help characterize the behavior of tellurium under severe light-water reactor (LWR) accident conditions. The release of tellurium from the fuel rods is dependent upon the rate and extent of cladding oxidation. Tellurium has been found to be considerably retained by metallic Zircaloy cladding at test temperatures up to 1975/sup 0/C. The results indicate that the tellurium is bound by the Zircaloy cladding as zirconium telluride, but once the available zirconium metal is oxidized by the steam, tellurium is released in favor of continued zirconium oxide formation. The collection behavior of the released tellurium indicates that it is probably released from the fuel rods as SnTe and CsTe, rather than as elemental tellurium.
Date: January 1, 1986
Creator: Collins, J.L.; Osborne, M.F. & Lorenz, R.A.
Partner: UNT Libraries Government Documents Department

Effects of time and other variables on fission product release rates

Description: The releases of krypton and cesium from highly irradiated LWR fuel have been examined in detail. The main interest has been the effect of time on the rate of release and the effects of heatup and cooldown cycles. The minute-by-minute release rates for fission product /sup 85/Kr from commercial fuel irradiated in the H.B. Robinson PWR are shown. The release rate, fraction per minute, is calculated in the same manner as release rates given in NUREG-0772; the fission gas, cesium, and iodine release rate curve from that report is also shown.
Date: January 1, 1986
Creator: Lorenz, R.A.; Osborne, M.F. & Collins, J.L.
Partner: UNT Libraries Government Documents Department

Fission product release from simulated LWR fuel. [Loss of coolant or spent fuel transportation accident conditions]

Description: A series of tests has been conducted with simulated LWR fuel as part of a program for determining the quantities and characteristics of radiologically significant fission products that can be released under postulated spent-fuel transportation accident (SFTA) conditions and successfully terminated loss-of-coolant accident (LOCA) conditions. These tests were performed in either flowing-steam or dry-air atmospheres with Zircaloy-4-clad fuel-rod segments that contained unirradiated UO/sub 2/ pellets coated with radioactively traced CsOH, CsI, and TeO/sub 2/. A summary of the test conditions and amounts released are given. Cesium release associated with the implanted CsOH appeared to be limited by the formation of low-volatility uranate compounds. Iodine release was observed primarily as CsI, but also as I/sub 2/; in addition, at test temperatures of 900/sup 0/C and above, significant migration of the CsI to the cooler ends of the fuel-rod segments was noted. Tellurium release was markedly restricted by rapid reaction with the Zircaloy cladding. The tests in air yielded enhanced releases of cesium and iodine, and considerable swelling of the oxidized UO/sub 2/. As anticipated, measured release fractions were greater when the test rods were ruptured at temperature by internal pressure than when the cladding failures were machined in the rods prior to testing.
Date: July 1, 1978
Creator: Lorenz, R.A.; Collins, J.L. & Manning, S.R.
Partner: UNT Libraries Government Documents Department

Modeling fission product release from ruptured LWR fuel rods

Description: The principal objectives of the fission product release program are to determine the quantity of radiologically significant fission products released from defected LWR fuel rods under accident conditions, identify their chemical and physical forms, and interpret the results for use as input to computer models of postulated transportation and loss-of-coolant accidents. Experimental work with flowing steam in the temperature range 500 to 1200/sup 0/C and with dry air at 500/sup 0/C and 700/sup 0/C has been completed. One series of tests, the Implant Test Series, employed simulated fission products which were coated on unirradiated UO/sub 2/ fuel pellets; a second series, the Low Burnup Fuel Test Series, used fuel capsules irradiated to 1000 MWd/MT at high heat rating (560 to 660 W/cm), and a third series of experiments, the High Burnup Test Series, used fuel irradiated to 30,000 MWd/MT in the H.B. Robinson reactor at low heat rating (175 to 320 W/cm). Sufficient analytical results have been obtained to permit the formulation of a preliminary empirical model for cesium release in steam. The model assumes that cesium release is the sum of two components: burst release (that carried out with escaping plenum gas when the rod ruptures) and diffusion release (that diffusing from the gap space after the plenum gas has vented).
Date: January 1, 1978
Creator: Lorenz, R.A.; Collins, J.L. & Malinauskas, A.P.
Partner: UNT Libraries Government Documents Department

Fission product source terms for the LWR loss-of-coolant accident

Description: The principal objectives of the fission product release program currently in progress at Oak Ridge National Laboratory are to determine the quantity of radiologically significant fission products released from defected light water reactor (LWR) fuel rods under accident conditions, identify their chemical and physical forms, and interpret the results for use as input to computer models of postulated spent fuel transportation accidents (SFTAs) and loss-of-coolant accidents (LOCAs). The purpose of this paper is to summarize the source term models, which have been developed for cesium and iodine by this program, and to demonstrate the application of the source term models to the analysis of cesium and iodine release during a Pressurized Water Reactor (PWR) LOCA.
Date: January 1, 1978
Creator: Lorenz, R.A.; Collins, J.L. & Malinauskas, A.P.
Partner: UNT Libraries Government Documents Department

Fission product release from fuel under LWR accident conditions

Description: Three tests have provided additional data on fission product release under LWR accident conditions in a temperature range (1400 to 2000/sup 0/C). In the release rate data are compared with curves from a recent NRC-sponsored review of available fission product release data. Although the iodine release in test HI-3 was inexplicably low, the other data points for Kr, I, and Cs fall reasonably close to the corresponding curve, thereby tending to verify the NRC review. The limited data for antimony and silver release fall below the curves. Results of spark source mass spectrometric analyses were in agreement with the gamma spectrometric results. Nonradioactive fission products such as Rb and Br appeared to behave like their chemical analogs Cs and I. Results suggest that Te, Ag, Sn, and Sb are released from the fuel in elemental form. Analysis of the cesium and iodine profiles in the thermal gradient tube indicates that iodine was deposited as CsT along with some other less volatile cesium compound. The cesium profiles and chemical reactivity indicate the presence of more than one cesium species.
Date: January 1, 1983
Creator: Osborne, M.F.; Lorenz, R.A.; Norwood, K.S.; Collins, J.L. & Wichner, R.P.
Partner: UNT Libraries Government Documents Department

Fission product release from highly irradiated LWR fuel

Description: A series of experiments was conducted with highly irradiated light-water reactor fuel rod segments to investigate fission products released in steam in the temperature range 500 to 1200/sup 0/C. (Two additional release tests were conducted in dry air.) The primary objectives were to quantify and characterize fission product release under conditions postulated for a spent-fuel transportation accident and for a successfully terminated loss-of-coolant accident (LOCA). In simulated, controlled LOCA-type tests, release at the time of rupture proved to be more significant than the diffusional release that followed. Comparison of the release data for the dry-air tests with the release data of similarly conducted tests in steam indicated significant increases in the releases of iodine, ruthenium, and cesium in air. Various parameters that affect fission product release are discussed, and experimental observations and analysis of the chemical behavior of releasable fission products in inert, steam, and dry-air atmospheres are examined.
Date: February 1, 1980
Creator: Lorenz, R.A.; Collins, J.L.; Malinauskas, A.P.; Kirkland, O.L. & Towns, R.L.
Partner: UNT Libraries Government Documents Department

Water washes and caustic leaches of sludge from Hanford Tank S-101 and water washes of sludge from Hanford Tank C-103

Description: In 1993, the Department of Energy (DOE) selected the enhanced sludge washing (ESW) process as the baseline for pretreatment of Hanford tank sludges. The ESW process uses a series of water washes and caustic leaches to separate nonradioactive components such as aluminum, chromium, and phosphate from the high-level waste sludges. If the ESW process is successful, the volume of immobilized high-level waste will be significantly reduced. The tests on the sludge from Hanford Tank S-101 focused on the effects of process variables such as sodium hydroxide concentration (1 and 3 M), temperature (70 and 95 C), and leaching time (5, 24, 72, and 168 h) on the efficacy of the ESW process with realistic liquid-to-solid ratios. Another goal of this study was to evaluate the effectiveness of water washes on a sludge sample from hanford Tank C-103. The final objective of this study was to test potential process control monitors during the water washes and caustic leaches with actual sludge. Both {sup 137}Cs activity and conductance were measured for each of the water washes and caustic leaches. Experimental procedures, a discussion of results, conclusions and recommendations are included in this report.
Date: July 1, 1998
Creator: Hunt, R.D.; Collins, J.L. & Chase, C.W.
Partner: UNT Libraries Government Documents Department

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

Description: One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.
Date: February 1, 2000
Creator: Lee, D.D. & Collins, J.L.
Partner: UNT Libraries Government Documents Department

Development and testing of inorganic sorbents made by the internal gelation process for radionuclide and heavy metal separations

Description: The objectives of this task are to develop, prepare, and test microspheres and granular forms of inorganic ion exchangers to remove radionuclides and heavy metals from waste streams occurring at various sites. Several inorganic materials, such as hexacyanoferrates, titanates, phosphates, and oxides have high selectivities and efficiencies for separating and removing radionuclides such as uranium, technetium, cesium, and strontium, and metals such as cobalt, silver, zinc, and zirconium from aqueous waste streams. However, these sorbents frequently exist only as powders and consequently are not readily adaptable to continuous processing such as column chromatography. Making these inorganic ion exchangers as microspheres or granular forms improves the flow dynamics for column operations and expands their practical applications. Microspheres of several materials have been prepared at ORNL, and the effectiveness of zirconium monohydrogen phosphate and hydrous titanium oxide microspheres for removing radionuclides from hot cell waste solutions has been demonstrated.
Date: November 29, 1995
Creator: Egan, B. Z.; Collins, J. L.; Anderson, K. K. & Chase, C. W.
Partner: UNT Libraries Government Documents Department

Aluminum removal from washed sludge

Description: Purpose of this project is to reduce the volume of storage tank sludge to be treated by removing the Al and other nonradioactive components. In initial sludge surrogate studies, Al, Cr, and Zn showed the highest solubility in NaOH solutions; Ce and Zr were the least soluble of the elements tested. Removal of Fe and Bi approached 2%, the rest of the elements studied showed <1% removal. Amount of Al removed increased as the NaOH conc. increased from 0.1 to 6 M. Sequential washing of the sludge surrogate with 3 M NaOH removed 84% of the Al, 39% of the Cr, and 65% of the Zn. Surrogate sludges containing U and Th were also studied.
Date: December 31, 1995
Creator: Egan, B.Z.; Collins, J.L. & Ensor, D.D.
Partner: UNT Libraries Government Documents Department

Development of inorganic ion exchangers for nuclear waste remediation. 1997 annual progress report

Description: 'In this research program, Oak Ridge National Laboratory (ORNL) is collaborating with Texas A and M University in the development of highly selective inorganic ion exchangers for the removal of cesium and strontium from nuclear tank-waste and from groundwater. Inorganic ion exchangers are developed and characterized at Texas A and M University; ORNL is involved in preparing the powders in engineered forms and testing the performance of the sorbents in actual nuclear waste solutions. The Texas A and M studies are divided into two main categories: (1) exchangers for tank wastes and (2) exchangers for groundwater remediation. These are subdivided into exchangers for use in acid and alkaline solutions for tank wastes and those that can be recycled for use in groundwater remediation. The exchangers will also be considered for in situ immobilization of radionuclides. The approach will involve a combination of exchanger synthesis, structural characterization, and ion exchange behavior. ORNL has developed a technique for preparing inorganic ion exchangers in the form of spherules by a gel-sphere internal gelation process. This technology, which was developed and used for making nuclear fuels, has the potential of greatly enhancing the usability of many other special inorganic materials because of the improved flow dynamics of the spherules. Also, pure inorganic spherules can be made without the use of binders. ORNL also has access to actual nuclear waste in the form of waste tank supernatant solutions for testing the capabilities of the sorbents for removing the cesium and strontium radionuclides from actual waste solutions. The ORNL collaboration will involve the preparation of the powdered ion exchangers, developed and synthesized at Texas A and M, in the form of spherules, and evaluating the performance of the exchangers in real nuclear waste solutions. Selected sorbents will be provided by Texas A and M for potential ...
Date: September 1997
Creator: Egan, B. Z.; Clearfield, A. & Collins, J. L.
Partner: UNT Libraries Government Documents Department

Characterization and chemistry of fission products released from LWR fuel under accident conditions

Description: Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000/sup 0/C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab.
Date: January 1, 1984
Creator: Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A. & Wichner, R.P.
Partner: UNT Libraries Government Documents Department