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COST OF BLENDING URANIUM OF TWO DIFFERENT U$sup 235$ ISOTOPIC ENRICHMENTS. Values of Uranium from the "USAEC Unclassified Pricing List"

Description: >Cost tables and instructions for calculating the cost of blending two different isotopic enrichments of U/sup 235/ to form an intermediate assay are presented. A sample calculation is included and information on reactor fuel cycles is given. (J. R. D.)
Date: October 14, 1958
Partner: UNT Libraries Government Documents Department

TABLES OF TWO U$sup 235$ FISSION SPECTRA

Description: The U/sup 235/ fission spectrum is tabulated as a function of energy for two analytic representations. The table contains the distribution functions, their first derivatives, and their first integrals up to 10 Mev. (auth)
Date: October 1, 1957
Creator: Howerton, R.J.; Bengston, J. & French, S.J.
Partner: UNT Libraries Government Documents Department

Neutron cross sections for uranium-235 (ENDF/B-VI release 3)

Description: The resonance parameters in ENDF6 (Release 2) U235 were adjusted to make the average capture and fission cross sections below 900 eV agree with selected differential capture and fission measurements. The measurements chosen were the higher of the credible capture measurements and the lower of the fission results, yielding a higher epithermal alpha. In addition, the 2,200 m/s cross sections were adjusted to obtain agreement with the integral value of K1. As a result, criticality calculations for thermal benchmarks, and agreement with a variety of integral parameters, are improved.
Date: May 1, 1996
Creator: Lubitz, C.
Partner: UNT Libraries Government Documents Department

Survey of available technology for auditing $sup 235$U enrichment in cascade equipment

Description: A search for possible nondestructive techniques which might be utilized by a safeguards inspection team for auditing /sup 235/U enrichment of UF/sub 6/ in gaseous diffusion cascade equipment without access to the UF/sub 6/ itself was carried out at the ORGDP. The work included a literature survey of safeguards techniques in use by others, and a search for other methods or equipment which might be applied to the present objective. Various possible methods are evaluated briefly. Probable limited access to cascade equipment precludes the use of methods involving bulky equipment, including neutron and gamma generators, coincident or anticoincident counters and high energy isotopic neutron sources of sufficient strength to produce a measurable fission rate in /sup 238/U which requires large shields. Irradiation with low energy neutrons followed by detection of fission events and passive gamma measurements could provide information on the /sup 235/U content, but not on the 2/sup 238/U. Two methods upon which preliminary investigations were carried out are considered to warrant further development. The first involves a direct measurement of the /sup 235/U gamma radiation from UF/sub 6/ within a pipe and a measurement of the absorption of gamma rays from an external gamma source when these rays pass through the pipe walls and UF/sub 6/. The /sup 235/U gamma measurement indicates the amount of this isotope present, and the gamma absorption measurement, when corrected for pipe-wall absorption, indicates the total uranium present. Experimental tests indicated this method to be well within the scope of current technology. The tests included an evaluation of a commercially available ultrasonic gauge which could be used for pipe-wall thickness measurements. The second method involves the irradiation of the UF/sub 6/ with a moderated neutron source and the detection of gamma rays resulting from neutron capture in /sup 238/U and fissions induced in ...
Date: November 1, 1973
Creator: Bailey, J.C.
Partner: UNT Libraries Government Documents Department

Nondestructive assay of fissile material samples in support of nuclear safeguards

Description: From nuclear science symposium; San Francisco, California, USA (14 Nov 1973). Samples of fissile material can be assayed by bombarding with 300- to 600- keV neutrons and counting delayed neutrons from fission. Interrogating neutron energy selection is based upon considerations of sample penetrability and insensitivity of response to nonfissile isotopes. Significant cost savings in nuclear safeguards and quality control are possible. (auth)
Date: January 1, 1973
Creator: Evans, A.E. Jr.
Partner: UNT Libraries Government Documents Department

Multilevel fitting of {sup 235}U resonance data sensitive to Bohr-and Brosa-fission channels

Description: The recent determination of the K, J dependence of the neutron induced fission cross section of {sup 235}U by the Dubna group has led to a renewed interest in the mechanism of fission from saddle to scission. The K quantum numbers designate the so-called Bohr fission channels, which describe the fission properties at the saddle point. Certain other fission properties, e.g., the fragment mass and kinetic-energy distribution, are related to the properties of the scission point. The neutron energy dependence of the fragment kinetic energies has been measured by Hambsch et al., who analyzed their data according to a channel description of Brosa et al. How these two channel descriptions, the saddle-point Bohr channels and the scission-point Brosa channels, relate to one another is an open question, and is the subject matter of the present paper. We use the correlation coefficient between various data sets, in which variations are reported from resonance to resonance, as a measure of both-the statistical reliability of the data and of the degree to which different scission variables relate to different Bohr channels. We have carried out an adjustment of the ENDF/B-VI multilevel evaluation of the fission cross section of {sup 235}U, one that provides a reasonably good fit to the energy dependence of the fission, capture, and total cross sections below 100 eV, and to the Bohr-channel structure deduced from an earlier measurement by Pattenden and Postma. We have also further explored the possibility of describing the data of Hambsch et al. in the Brosa-channel framework with the same set of fission-width vectors, only in a different reference system. While this approach shows promise, it is clear that better data are also needed for the neutron energy variation of the scission-point variables.
Date: May 1, 1995
Creator: Moore, M.S.
Partner: UNT Libraries Government Documents Department

Development of DU-AGG (Depleted Uranium Aggregate)

Description: Depleted uranium oxide (UO{sub 2} or U0{sub 3}) powder was mixed with fine mineral additives, pressed, and heated to about 1,250{degree}C. The additives were chemically constituted to result in an iron-enriched basalt (IEB). Melting and wetting of the IEB phase caused the urania powder compact to densify (sinter) via a liquid phase sintering mechanism. An inorganic lubricant was found to aid in green-forming of the body. Sintering was successful in oxidizing (air), inert (argon), or reducing (dry hydrogen containing) atmospheres. The use of ground U0{sub 3} powders (93 vol %) followed by sintering in a dry hydrogen-containing atmosphere significantly increased the density of samples (bulk density of 8.40 g/cm{sup 3} and apparent density of 9.48 g/cm{sup 3}, open porosity of 11.43%). An improvement in the microstructure (reduction in open porosity) was achieved when the vol % of U0{sub 3} was decreased to 80%. The bulk density increased to 8.59 g/cm{sup 3}, the apparent density decreased slightly to 8.82 g/cm{sup 3} (due to increase of low density IEB content), while the open porosity decreased to an excellent number of 2.78%. A representative sample derived from 80 vol % U0{sub 3} showed that most pores were closed pores and that, overall, the sample achieved the excellent relative density value of 94.1% of the estimated theoretical density (composite of U0{sub 2} and IEB). It is expected that ground powders of U0{sub 3} could be successfully used to mass produce lowcost aggregate using the green-forming technique of briquetting.
Date: September 1, 1995
Creator: Lessing, P.A.
Partner: UNT Libraries Government Documents Department

ANALYSIS OF THE SECOND ZEUS CRITICAL EXPERIMENT

Description: The Zeus experiments have been designed to test the adequacy of {sup 235}U cross sections in the intermediate energy range. The first Zeus experiment achieved initial criticality on April 26, 1999 with 10 HEU platters and 79 platters of graphite. The second experiment reached a critical condition on October 24, 2000, with 9 HEU platters and 54 platters of graphite. The first experiment, which has been described and analyzed previously, had a C/{sup 235}U ratio of approximately 52:1. The corresponding ratio for the second experiment was approximately 40:1. This summary describes the analysis of the second Zeus critical experiment.
Date: January 1, 2001
Creator: MOSTELLER, R. & JAEGERS, P.
Partner: UNT Libraries Government Documents Department

Statistical properties of the S-wave resonances of {sup 235}U

Description: The resonance parameters of {sup 235}U in the energy range 0 eV to 2.25 keV were obtained from a generalized least squares analysis of a large set of experimental data using the Reich-Moore formalism in the fitting code SAMMY. The aim of the present paper is to present the statistical properties of the s-wave resonance parameters generated from this study.
Date: June 1, 1997
Creator: Leal, L.C.; Derrien, H. & Larson, N.M.
Partner: UNT Libraries Government Documents Department

VARIATIONS IN ISOTOPIC CONTENT OF NATURAL URANIUM

Description: Uranium ore concentrates from seventeen world sources were compared to a standard to determine variations in isotopic content. A spread of about 0.06% in U/sup 235/ content was indicated for the concentrates analyzed. Domestic sources showed much wider variations than those from other parts of the world. (auth)
Date: June 26, 1961
Creator: Smith, R.F.; Eby, R.E. & Turok, C.W.
Partner: UNT Libraries Government Documents Department

Filter Paper: Solution to High Self-Attenuation Corrections in HEPA Filter Measurements

Description: An 8 by 8 by 6 inch High Efficiency Particulate Air (HEPA) filter was measured as part of a uranium holdup survey in June of 2005 as it has been routinely measured every two months since 1998. Although the survey relies on gross gamma count measurements, this was one of a few measurements that had been converted to a quantitative measurement in 1998. The measurement was analyzed using the traditional Generalized Geometry Holdup (GGH) approach, using HMS3 software, with an area calibration and self-attenuation corrected with an empirical correction factor of 1.06. A result of 172 grams of {sup 235}U was reported. The actual quantity of {sup 235}U in the filter was approximately 1700g. Because of this unusually large discrepancy, the measurement of HEPA filters will be discussed. Various techniques for measuring HEPA filters will be described using the measurement of a 24 by 24 by 12 inch HEPA filter as an example. A new method to correct for self attenuation will be proposed for this measurement Following the discussion of the 24 by 24 by 12 inch HEPA filter, the measurement of the 8 by 8 by 6 inch will be discussed in detail.
Date: October 1, 2005
Creator: Oberer, R.B.; Harold, N.B.; Gunn, C.A.; Brummett, M. & Chaing, L.G.
Partner: UNT Libraries Government Documents Department

Withdrawal assay monitoring at US Enrichment Facilities

Description: The United States Enrichment Corporation (USEC) controls two uranium enrichment facilities that produce enriched uranium for both military and commercial use. The process requires both feed and withdrawal operations. The withdrawal process requires both product (enriched uranium) withdrawal stations and tails (depleted uranium) withdrawal stations. A previous prototype system, ``X-330 Tails Cylinder Assay Monitor,`` was developed as a demonstration for the tails withdrawal station at the Portsmouth Gaseous Diffusion Plant (PORTS). The prototype system was done in response to potential problems with the original method for determining the hourly weighted assay averages that are used to calculate the final weighted assay of the cylinder. In the original method the {sup 235}U assay of uranium hexaflouride withdrawn from PORTS cascade into tails cylinders is determined every 5 min by measurements from an in-line assay mass spectrometer. An average value for a 1-h period is then calculated by area control room personnel and assigned to the accumulated weight in the cylinder for the period. A potential problem with this method is that cylinder weight is not automatically recorded as often as the assay. The assay and withdrawal rate can both vary during the given period. This variation results in inaccuracies in the hourly weighted assays that are used to calculate the final weighted assay of the cylinder. Laboratory analysis is considered to be the most accurate method for determining the final cylinder assay; however, the cost and safety considerations of redundant cylinder handling limit the number of cylinders sampled to less than 10%.
Date: January 1996
Creator: Smith, D. E.
Partner: UNT Libraries Government Documents Department

Waste Processing To Support {sup 99}Mo Production at Sandia National Laboratories

Description: As part of the Isotope Production Program at Sandia National Laboratories New Mexico (SNL/NM), procedures are being finalized for the production of {sup 99}Mo from the irradiation of {sup 235}U-coated stainless steel targets at the Technical Area (TA) V reactor and hot cell facilities. Methods have been identified and tested for the management of the non-product (waste) material as the final step in the production process. These methods were developed utilizing the waste material from a series of cold and hot tests, beginning with depleted uranium powder and culminating with a test involving an irradiated {sup 235}U target with an initial fission product inventory of approximately 18,000 Ci at the end of the irradiation cycle.
Date: June 1, 1997
Creator: Longley, Susan; Carson, Susan & McDonald, Marion
Partner: UNT Libraries Government Documents Department

Calculation of 1.25% 235U enriched UO2 solution safe slab, safe cylinder diameter, minimum safe mass, and ion exchange module for the CVDF

Description: Support calculations were performed to establish safe parameters such as fissionable material slab thickness, diameter and safe mass. These calculations were performed by MCNP for the balance of plant equipment that contains homogeneous UO{sub 2} solutions with a maximum enrichment of 1.25 Wt% {sup 235}U . The calculations were performed with the most limiting concentration of moderator and reflection so that only the safety parameters identified in the problem description need to be controlled. These calculations represent the most limiting cases for all uranium enrichments and transuranic levels due to fuel exposure for balance of plant equipment used for handling of waste water containing fissionable materials from the MCO draining and drying activities.
Date: June 26, 1997
Creator: Roblyer, S.P.
Partner: UNT Libraries Government Documents Department

Method for fabricating {sup 99}Mo production targets using low enriched uranium, {sup 99}Mo production targets comprising low enriched uranium

Description: A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material (low enriched U) circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.
Date: December 31, 1993
Creator: Wiencek, T.C.; Matos, J.E. & Hofman, G.L.
Partner: UNT Libraries Government Documents Department

Limitations on the precision of 238U/235U measurements and implications for environmental monitoring

Description: The ability to determine the isotopic composition of uranium in environmental samples is an important component of the International Atomic Energy Agency`s (IAEA) safeguards program, and variations in the isotopic ratio {sup 238}U/{sup 235}U provide the most direct evidence of isotopic enrichment activities. The interpretation of observed variations in {sup 238}U/{sup 235}U depends on the ability to distinguish enrichment from instrumental biases and any variations occurring in the environment but not related to enrichment activities. Instrumental biases that have historically limited the accuracy of {sup 238}U/{sup 235}U determinations can be eliminated by the use of the {sup 233}U/{sup 236}U double-spike technique. With this technique, it is possible to determine the {sup 238}U/{sup 235}U in samples to an accuracy equal to the precision of the measurement, ca. 0.1% for a few 10`s of nanograms of uranium. Given an accurate determination of {sup 238}U/{sup 235}U, positive identification of enrichment activities depends on the observed value being outside the range of {sup 238}U/{sup 235}U`s expected as a result of natural or environmental variations. Analyses of a suite of soil samples showed no variation beyond 0.2% in {sup 238}U/{sup 235}U.
Date: August 1, 1997
Creator: Russ III, G.P.
Partner: UNT Libraries Government Documents Department

AVLIS Criticality risk assessment

Description: Evaluation of criticality safety has become an important task in preparing for the Atomic Vapor Laser Isotope Separation (AVLIS) uranium enrichment runs that will take place during the Integrated Process Demonstration (IPD) at Lawrence Livermore National Laboratory (LLNL). This integrated operation of AVLIS systems under plant-like conditions will be used to verify the performance of process equipment and to demonstrate the sustained integrated enrichment performance of these systems using operating parameters that are similar to production plant specifications. Because of the potential criticality concerns associated with enriched uranium, substantial effort has been aimed towards understanding the potential system failures of interest from a criticality standpoint, and evaluating them in detail. The AVLIS process is based on selective photoionization of uranium atoms of atomic weight 235 (U-235) in a vapor stream, followed by electrostatic extraction. The process is illustrated in Figure 1. Two major subsystems are involved: the uranium separator and the laser system. In the separator, metallic uranium is fed into a crucible where it is heated and vaporized by an electron beam. The atomic U-235/U-238 vapor stream moves away from the molten uranium and is illuminated by precisely tuned beams of dye laser light. Upon absorption of the tuned dye laser light, the U-235 atoms become excited and eject electrons (become photoionized), giving them a net positive charge. The ions of U-235 are moved preferentially by an electrostatic field to condense on the product collector, forming the enriched uranium product. The remaining vapor, which is depleted in U-235 (tails), passes unaffected through the photoionization/extractor zone and accumulates on collectors in the top of the separator. Tails and product collector surfaces operate at elevated temperatures so that deposited materials flow as segregated liquid streams. The separated uranium condensates (uranium enriched in U-235 and uranium depleted in U-235) are cooled and ...
Date: April 29, 1998
Creator: Brereton, S. J.
Partner: UNT Libraries Government Documents Department