The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Disposition Projects (FDP) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control & Accountability, and to meet criticality safety controls. This report covers calibration of the detectors in order to support holdup measurements in the C and D out-gassing ovens. These ovens were used to remove gas entrained in billet assembly material prior to the billets being extruded into rods by the extrusion press. A portable high purity germanium detection system was used to determine highly enriched uranium (HEU) holdup and to determine holdup of 235U, 237Np, and 241Am that were observed in these components. The detector system was run by an EG&G Dart system that contains the high voltage power supply and signal processing electronics. A personal computer with Gamma-Vision software was used to control the Dart MCA and provide space to store and manipulate multiple 4096-channel -ray spectra. The measured 237Np and 241Am contents were especially important in these components because their presence is unusual and unexpected in 321-M. It was important to obtain a measured value of these two components to disposition the out-gassing ovens and to determine whether a separate waste stream was necessary for release of these contaminated components to the E-Area Solid Waste Vault. This report presents determination of the calibration constants from first principles for determination of 241Am and 237Np using this detection system and compares the values obtained for 237Np with the calibration factors obtained with a subsequent measurement using a point source of radioactive equilibrium 237Np/233Pa.
This paper describes development work to prepare a method to measure absolute239Pu content and Pu-isotopics by ICP-MS in acidified Hanford DOE-site samples which are very high in90Sr,99Tc, and137Cs radioactivity and which are frequently high in organic carbon content. Samples with very large90Sr and137Cs contents have historically been difficult to analyze for Pu content by each of three alpha-counting techniques in use at
SRS, and analysis by ICP-MS in these samples is complicated by the high organics content. We report an ion exchange chemical
preparation to obtain fraction of Pu that does not contain any fission product contribution and no interfering organics to
allow measure of absolute239Pu and of239Pu through241Pu isotopics by ICP-MS. The method uses a242Pu spike to measure Pu recovery and is demonstrated in this paper with three distinct commercially available resins and with
over 300 samples. Measured absolute239Pu contents in sixty-three spiked/unspiked duplicates have agreed within 15% precision. Overall242Pu recoveries were near 90% with 25% precision. Comparisons of absolute239Pu contents measured directly on three samples agreed within the quoted 25% uncertainty.
The Savannah River Site (SRS) Burial Ground had a container labeled as Box 33 for which they had no reliable solid waste stream designation. The container consisted of an outer box of dimensions 48″ × 46″×66″
and an inner box that contained high density and high radiation dose material. From the outer box Radiation Control measured
an extremity dose rate of 22 mrem/h. With the lid removed from the outer box, the maximum dose rate measured from the inner
box was 100 mrem/h extremity and 80 mrem/h whole body. From the outer box the material was sufficiently high in density that
the Solid Waste Management operators were unable to obtain a Co-60 radiograph of the contents. Solid Waste Management requested
that the Analytical Development Section of Savannah River National Laboratory perform a γ-ray assay of the item to evaluate
the radioactive content and possibly to designate a solid waste stream. This paper contains the results of three models used
to analyze the measured γ-ray data acquired in an unusual configuration.
We describe tests of EiChrom Industries' Ni-selective ion exchange resin for use in analysis of63Ni in Savannah River Site high level waste. We report measurement of63Ni content in two sets of Savannah River Site glass product from the Defense Waste Processing Facility. The63Ni β-decay activity was chemically separated in quintuplicate from fission product and plutonium α-β activities of up to 103 times the observed63Ni content. The separation used a Ni-dimethlyglyoxime precipitation followed by radiochemical purification using the Ni-selective
extraction chromatography resin. Further removal of interfering activity was based on diagnosis of observed radioactivity
in each successive product phase. We analyzed eleven plant glass product samples using seven spiked standard addition duplicates
to measure63Ni recovery in the separations. Selected liquid scintillation β-decay spectra are shown to validate the method. Interpretation
of accuracy is based upon three distinct comparisions to predicted63Ni content.
We have developed a chemical separation technique that allows the radiochemical determination of the californium -decay content in Department of Energy (DOE) high level wastes from the Hanford and Savannah River sites. The chemical separation technique uses a series of column extraction chromatography steps that use Eichrom Industries' lanthanide and actinide +3 oxidation state selective Ln-resin® and the transuranic selective +4 oxidation state TRU-resin® to obtain intermediate product phases in dilute nitric acid. The technique has been demonstrated on three types of authentic DOE high and low level waste samples. We obtain discrimination from Pu -activity by a factor of over 200 and from 244Cm -activity by a factor approaching 1700. Californium recoveries are measured by addition of a 249Cf spike and are in the range of 50% to 90% in the synthetic samples and are in the range of 1.4% to 48% for the authentic DOE waste samples.
In this paper we describe use of the Aquila active well neutron coincidence counter for nuclear material assays of 235U in multiple analytical techniques at Savannah River Site (SRS), at the Savannah River National Laboratory (SRNL), and at
Argonne West National Laboratory (AWNL). The uses include as a portable passive neutron counter for field measurements searching
for evidence of 252Cf deposits and storage; as a portable active neutron counter using an external activation source for field measurements searching
for trace 235U deposits and holdup; for verification measurements of U-Al reactor fuel elements; for verification measurements of uranium
metal; and for verification measurements of process waste of impure uranium in a challenging cement matrix. The wide variety
of uses described demonstrate utility of the technique for neutron coincidence verification measurements over the dynamic
ranges of 100–5000 g for U metal, 200–1300 g for U-Al, and 8–35 g for process waste. In addition to demonstrating use of the
instrument in both the passive and active modes, we also demonstrate its use in both the fast and thermal neutron modes.
Gamma-ray holdup measurements of a Mossbauer spectroscopy instrument are described and modeled. In the qualitative acquisitions
obtained in a low background area of Savannah River National Laboratory, only Am-241 and Np-237 activity were observed. The
Am-241 was known to be the instrumental activation source, while the Np-237 is clearly observed as a source of contamination
internal to the instrument. The two sources of activity are modeled separately in two acquisition configurations using two
separate modeling tools. The results agree well, demonstrating a content of (1980 ± 150) μCi Am-241 and (110 ± 50) μCi of
We report the first observation of confirmed 79Se activity in Savannah River Site high level fission product waste. 79Se was measured after a seven step chemical treatment to remove interfering activity from 137Cs, 90Sr, and plutonium at levels 105 times higher than the observed 79Se content and to remove 99Tc at levels 300 times higher than observed 79Se. 79Se was measured by liquid scintillation -decay counting after specific tests to eliminate uncertainties from possible contributions from 99Tc, 147Pm, 151Sm, 93Zr, or 241Pu, whose -decay spectra could appear similar to that of 79Se, and whose content would be expected at levels near or greater than 79Se.
Authors:R. Dewberry, V. Casella, R. Sigg, and N. Bhatt
This paper represents a description of eight compiled benchmark tests conducted to probe and to demonstrate the extensive
utility of the Ortec ISOTOPIC gamma-ray analysis software program. The paper describes tests of the programs capability to
perform finite geometry correction factors and sample-matrix-container photon absorption correction factors. Favorable results
are obtained in all benchmark tests.
Authors:W. Kinard, N. Bibler, C. Coleman, and R. Dewberry
Highly radioactive waste from defense-related activities at the Savannah River Site in South Carolina are to be incorporated into a borosilicate glass in the Defense Waste Processing Facility (DWPF) for long-tem geological isolation. Processing and repository safety considerations require the determination of 24 radioisotopes that meet the reporting criteria. These isotopes include fission products, activation products, and daughter nuclei that grow into the waste. Four isotopes,137Cs,90Sr,238Pu and238U will be routinely measured in the DWPF operation for process control. This work shows that the concentrations of the other 20 reportable radioisotopes in the final glass product can be predicted from a thorough characterization of the high level waste (HLW) tanks and a knowledge of the concentrations of the major non-radioactive components in the vitrification process.