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Predicting 232U Content in Uranium

Description: The minor isotope 232U may ultimately be used for detection or confirmation of uranium in a variety of applications. The primary advantage of 232 U as an indicator of the presence of enriched uranium is the plentiful and penetrating nature of the radiation emitted by its daughter radionuclide 208Tl. A possible drawback to measuring uranium via 232U is the relatively high uncertainty in 232U abundance both within and between material populations. An important step in assessing this problem is to ascertain what determines the 232U concentration within any particular sample of uranium. To this end, we here analyze the production and eventual enrichment of 232 U during fuel-cycle operations. The goal of this analysis is to allow approximate prediction of 232 U quantities, or at least some interpretation of the results of 232U measurements. We have found that 232U is produced via a number of pathways during reactor irradiation of uranium and is subsequently concentrated during the later enrichment of the uranium' s 235U Content. While exact calculations are nearly impossible for both the reactor-production and cascade-enrichment parts of the prediction problem, estimates and physical bounds can be provided as listed below and detailed within the body of the report. Even if precise calculations for the irradiation and enrichment were possible, the ultimate 212U concentration would still depend upon the detailed fuel-cycle history. Assuming that a thennal-diffusion cascade is used to produce highly enriched uranium (HEU), dilution of reactor-processed fuel at the cascade input and the long-term holdup of 232U within the cascade both affect the 232U concentration in the product. Similar issues could be expected to apply for the other isotope-separation technologies that are used in other countries. Results of this analysis are listed below: 0 The 232U concentration depends strongly on the uranium enrichment, with depleted uranium (DU) ...
Date: January 7, 1999
Creator: Peurrung, AJ
Partner: UNT Libraries Government Documents Department

New {sup 232}U/{sup 228}Th Gamma Source for Tritiated Water Monitor

Description: The {sup 232}U/{sup 228}Th source has been tested for operation with the tritiated water monitor. This source has an effective half-life of 73 years and is an attractive replacement for the reactor activated {sup 24}Na sources, which have half-life of only 15 hours. The tests described in this report appraise the adequacy of this source by comparing its performance to that of {sup 24}Na used in earlier studies. The new source has already been used successfully to assist recovery operations, and it is now apparent that the monitor is ready to be turned over to SRP for routine use.
Date: July 17, 2001
Creator: Winn, W.G.
Partner: UNT Libraries Government Documents Department

Radiation Hazards From Recycled Reactor Fuel

Description: The radiation hazards associated with recycled nuclear reactor fuels will greatly complicate the handling and refabrication of these fuels. This problem is most serious with U/sup 233/ and Pu fuels where the presence of U/sup 232/ and thue heavier isotopes of Pu contribute energetic alpha, gamma, and neutron radiations at levels many times that from isotopically pure U/sup 233/ and Pu/sup 239/. Present knowledge of the radiation hazards associated with recycled fuel and the additional data needed to make a thorough evaluation of these hazards are summarized. (auth)
Date: June 1, 1959
Creator: Arnold, E. D.
Partner: UNT Libraries Government Documents Department

A U-235 confirmation method with a discriminating view

Description: Reprocessed spent nuclear fuel that has gone through various stages of chemical processing is currently stored at the INEEL. The material consists of three categories: clean U-235 with radiation readings in the mR/h range, U-235 and fission products up to 900 mR/h, and U-235 with U-232 from 100--200 mR/h. The material is contained in plastic bottles and stored in steel structures consisting of seven vertically arranged individual compartments. A total of forty stacks reside in individual concrete wells. This material is considered hard to measure due to excessive radiation exposure to personnel involved with handling the material during mass and NaI U-235 confirmation measurements for Safeguards inventory purposes. A U-235 confirmation method was developed to assay the individual items in place with the ability to discriminate one item from the other items in the stack. Equipment used with this method includes a portable high-resolution gamma-ray detection system, an appropriate tungsten shield and collimator, and a laser-positioning device. A discrimination control test was incorporated to compare the gamma-ray signal of an item in place to the background signal when the item is removed. Total discrimination of the 186-keV gamma ray signal was achieved.
Date: September 1, 1998
Creator: McLaughlin, G.D.; Hartwell, J.K. & Reed, B.M.
Partner: UNT Libraries Government Documents Department

Enrichment Monitor for 235U Fuel Tubes

Description: This report describes the performance of this prototype y-monitor of 235 Uranium enrichment. In this proposed method y-rates associated with 235U and 232U are correlated with enrichment. Instrumentation for appraising fuel tubes with this method has been assembled and tested.
Date: August 22, 2001
Creator: Winn, W.G.
Partner: UNT Libraries Government Documents Department

A Tritiated-Water Detector with U-232/Th-228 Source

Description: The detection capabilities of the new U-232/Th-228 source are comparable to those of the Na-24 source. The main benefit in using the new source is the ease of operation. Elimination of the neutron activation step required for Na-24 sources saves about 24 hours in planning, scheduling, and executing. With the new U-232/Th-228 source, the monitor can be put in operation in less than 15 minutes. The long half-life of the U-232/Th-228 source also eliminates the need to record calibration and measurement times, as required for decay corrections when using a Na-24 source.
Date: May 29, 2001
Creator: Baumann, N.P.
Partner: UNT Libraries Government Documents Department

Radon Measurements at the Molten Salt Reactor Experiment (MSRE) Facility from August 1997 through April 1998

Description: Planned remediation at the Oak Ridge National Laboratory (ORNL) Molten Salt Reactor Experiment (MSRE) facility created the need to measure 220Rn and its progeny in air within work areas. Most of the original 233U fuel material is still present in the process system at the facility and will eventually be removed as decommissioning progresses. A hazard associated with the 233U material is the production of 220Rn from the 232U decay chain. Although 232U is in 233U material at a small mass fraction, it can equate to a significant activity level due to its short half-life. To illustrate the magnitudes of activity expected, consider the fuel material in the MSRE facility, which contains 232U at a mass concentration of 160 ppm. After a period of about 13 years, the 228Th activity would have equilibrated to about 3.5 Ci/kg of uranium. A kilogram of uranium would therefore be expected to produce 220Rn at a rate of 1.3E11 atoms/s, or 4.4E-2 Ci/s.
Date: June 1999
Creator: Coleman, R.L.
Partner: UNT Libraries Government Documents Department

ALPHA-GAMMA ANGULAR CORRELATION MEASUREMENTS WITH LIQUID SOURCES

Description: Alpha-gamma angular correlation measurements were made with solid sources of Am/sup 243/ and with liquid sources containing either Am/sup 243/ or an even-even alpha emitter in dilute perchloric acid solutions. Even-even alpha emitters studied are U/sup 232/, Th/sup 230/, and Ra/sup 226/ . Thicknesses of the soli d sources were controlled so that the neptunium recoils from one source were stopped in Am/sub 2/O/sub 3/, while recoils from the other sources were stopped in the aluminum, gold, or mica backing on which the sources were vaporized. The liquid sources were films consisting of 3 microliters of solution placed between a rubber hydrochloride membrane and a microscope cover glass, 1 cm/sup 2/ in circular cross section. The perchloric acid concentration of the liquid sources ranged from 0.5 to 3.0 molar. All of the angular correlations obtained with solid Am/sup 243/ sources were attenuated, the average attenuation coefficients being 0.29 450 deg C in a 0.01 for sources in which recoils were stopped in Am/sub 2/O/sub 3/, 0.20 450 deg C in a 0.01 for sources in which recoils were stopped in mica, 0.52 450 deg C in a 0.02 for sources in which recoils were stopped in gold, and 0.67 450 deg C in a 0.01 for sources in which recoils were stopped in aluminum. Unattenuated angular correlations were obtained with liquid sources containing Am/sup 243/ in 0.5 M and 1.0 M HClO/sub 4/ . For liquid sources containing Am/sup 243/ in 3.0 M HClO/sub 4/, the correlation was attenuated, with an average attenuation coefficient of 0.86 450 deg C in a 0.01. Attenuated angular correlations were also found with liquid sources containing an even-even nuclide in dilute aqueous solutions The average attenuation coefficients for the even-even nuclide liquid sources were G/sub 2/ = 0.75 450 deg C in a 0.05 ...
Date: February 1, 1963
Creator: Murphy, E.S. Jr.
Partner: UNT Libraries Government Documents Department

RATIO OF U-232 TO U-233 PRODUCED IN THE TBR

Description: The production of U/sup 232/ in the cost optimized TBR and oneregion thorlum oxide-uranium oxide slurry reactor has been estimated to be, respectively, 40 and 260 parts per million parts of U/sup 233/ produced. These production rates are compared with corresponding production rates in irradiated thorium rods and found to be comparable. Recommendations are made for increasing the purity of the U/sup 233/ product. (auth)
Date: May 23, 1955
Creator: Dresner, L
Partner: UNT Libraries Government Documents Department

Uranium Density and Enrichment in Fuel Tubes Determined from 232U and 235U Y-Activities

Description: Gamma spectroscopy is used to determine 235U density and enrichment in U-Al fuel tubes containing recycled fuel. A collimated HPGe Y-detector views the tube surface, such that U-Al disk volumes of 6.35 mm diameter and approximately 1.0 mm thickness are examined. The Y-activities from 232U and 235U, along with the tube design parameters, are used to deduce the attenuation-corrected results. Respective density and enrichment variations of less than 1 percent and less than 0.6e percent were measurable with 2000 sec counting time per tube location. Such measurements are useful for certifying tube quality and characterizing problems associated with blending the U-Al alloy.
Date: May 25, 1984
Creator: Winn, W.G.
Partner: UNT Libraries Government Documents Department

Fission, total and neutron capture cross section measurements at ORELA for {sup 233}U, {sup 27}Al and natural chlorine

Description: The authors have made use of the Oak Ridge Electron Linear Accelerator (ORELA) to measure the fission cross section of {sup 233}U in the neutron energy range of 0.36 eV to {approximately} 700 keV. This paper reports integral data and average cross sections. In addition they measured the total neutron cross section of {sup 27}Al and natural chlorine, as well as the capture cross section of Al over an energy range from 100 eV up to about 400 keV.
Date: August 1, 1998
Creator: Guber, K.H.; Spencer, R.R.; Leal, L.C.; Larson, D.C.; Santos, G. Dos; Harvey, J.A. et al.
Partner: UNT Libraries Government Documents Department

Actinides at the crossroads: ICP-MS or alpha spectrometry?

Description: The report contains viewgraphs only that summarize the following: Why turn to mass spectrometry for radiochemical measurements; What might be some advantages of using ICP mass spectrometry; Sensitivity of ETV-ICP-MS relative to decay counting (versus half-life); ICP-MS instrument detection limits for dissolved actinide isotopes; Effect of dissolved solids on USN-ICP-MS analysis; Polyatomic ion interferences in ICP-MS actinide measurements; Effect of operating conditions on uranium and protonated uranium signal; ICP mass spectrometry performance in actinide determinations; Determination of actinide elements in soil; Leachable Th-230 and Pu-239 in soil as determined by ICP-MS and alpha spectrometry; Leachable U-234 and U-238 in soil by ICP-MS and alpha spectrometry; Determination of uranium isotopic composition on smears; Activity ratios (U-234/U-238) as determined by mass spectrometry and alpha spectrometry; Uranium isotopic abundances as determined by TIMS and ICP-MS; and Comparison of uranium atom percentages determined by TIMS and ICP-MS. It is concluded that isotope dilution and radiochemical preparative techniques work well in radioanalytical applications of ICP-MS; radioanalytical ICP-MS data are equivalent to data from standard methods (TIMS, alpha spectrometry); and applications in radiation protection and earth sciences are certain to expand further.
Date: December 31, 1995
Creator: Crain, J.S.; Yaeger, J.S.; Smith, F.P.; Alvarado, J.A.; Smith, L.L.; Kiely, J.T. et al.
Partner: UNT Libraries Government Documents Department

Project SAPPHIRE uranium-beryllium dose rate analysis

Description: During a six-week period in the fall of 1994 a team of 31 US government and Y-12 personnel packaged and removed several thousand kilograms of material containing highly enriched uranium from the (former Soviet Union) Republic of Kazakhstan for interim storage at the Y-12 Plant in Oak Ridge, Tennessee. This classified mission, known as PROJECT SAPPHIRE, had been initiated at the request of the Kazakhstan government in order to rid itself of possible security problems. Planning for the mission included assurance of the health and safety of the team members, as well as compliance with all local, IAEA, and US government regulations regarding the handling, packaging, transportation, and storage of radioactive and fissile material. The mission classification restrictions were relaxed following the return of the team and material to the United States. The material to be removed, in the form of small billets and rods of uranium metal and uranium-beryllium alloy and oxide powder, was sealed by team members on site into two-liter steel cans. Two or three cans each were loaded into more than 400 IAEA certified fissile material shipping container, and each container was packed into a large steel drum for transport by US Air Force cargo planes to the United States.
Date: June 21, 1995
Creator: Cramer, S.N.; Lewis, K.D. & Moses, S.D.
Partner: UNT Libraries Government Documents Department