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.
NAA using 252Cf is used to address important areas of applied interest of the Savanah River Site (SRS). Sensitivity needs for many of the
applications are not severe; analyses are accomplished using a 21 mg 252Cf NAA facility. Because NAA allows analysis of bulk samples, it offers strong advantages for samples in difficult-to-digest
matrices when its sensitivity is sufficient. Following radiochemical separation with stable carrier addition, chemical yields
for a number of methods are determined by neutron activation of the stable carrier. In some of the cases where no suitable
stable carriers exist, the source has been used to generate radioactive tracers to yield separations.
Neutron activation analysis using low-flux isotopic neutron sources is put to use in addressing areas of applied interest in managing the Savannah River Site. Some of the applications are unique due the site's operating history and its chemical processing facilities. Because sensitivity needs for many of the applications are not severe, they can be accomplished using a 6 mg 252Cf neutron activation analysis facility. Overviews of the facility and several example applications are presented.
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:S. Howell, R. Sigg, F. Moore, and T. DeVol
A prompt gamma-ray neutron activation analysis (PGNAA) system was used to calibrate and validate a Monte Carlo model as a proof of principle for the quantification of chlorine in soil. First, the response of an n-type HPGe detector to point sources of 60Co and 152Eu was determined experimentally and used to calibrate an MCNP4a model of the detector. The refined MCNP4a detector model can predict the absolute peak detection efficiency within 12% in the energy range of 120–1400 keV. Second, a PGNAA system consisting of a light-water moderated 252Cf (1.06 g) neutron source, and the shielded and collimated HPGe detector was used to collect prompt gamma-ray spectra from Savannah River Site (SRS) soil spiked with chlorine. The spectra were used to calculate the minimum detectable concentration (MDC) of chlorine and the prompt gamma-ray detection probability. Using the 252Cf based PGNAA system, the MDC for Cl in the SRS soil is 4400 g/g for an 1800-second irradiation based on the analysis of the 6110 keV prompt gamma-ray. MCNP4a was used to predict the PGNAA detection probability, which was accomplished by modeling the neutron and gamma-ray transport components separately. In the energy range of 788 to 6110 keV, the MCNP4a predictions of the prompt gamma-ray detection probability were generally within 60% of the experimental value, thus validating the Monte Carlo model.
Authors:D. P. DiPrete, C. C. DiPrete, and R. A. Sigg
Waste cleanup efforts currently underway at the Savannah River Site have created a need to characterize 99Tc in the various high activity waste matrices currently in Site inventories. The traditional method our laboratory used for analyzing 99Tc in higher activity matrices was a solvent-solvent extraction method using Aliquat-336 in xylene. In an effort to eliminate the resulting generation of mixed wastes, a variety of different separation methodologies have been studied. Eichrom TEVA solid phase extractions using column technology have been employed in a case by case basis over the last several years. More recently, applications using Eichrom TEVA extraction discs and 3M Empore Tc extraction discs have also been explored.
Authors:R. Dewberry, V. Casella, R. Sigg, S. Salaymeh, F. Moore, and D. Pak
Visual Examination (VE) gloveboxes are used to remediate transuranic waste (TRU) drums at three separate facilities at the
Savannah River Site. Noncompliant items are removed before the drums undergo further characterization in preparation for shipment
to the Waste Isolation Pilot Plant (WIPP). Maintaining the flow of drums through the remediation process is critical to the
program’s seven-days-per-week operation. Conservative assumptions are used to ensure that glovebox contamination from this
continual operation is below acceptable limits. Holdup measurements using cooled HPGe spectrometers are performed in order
to confirm that these assumptions are conservative. 239Pu is the main nuclide of interest; however, 241Pu, equilibrium 237Np/233Pa and 238Pu (if detected) are typically assayed. At the Savannah River National Laboratory (SRNL) facility 243,244,245Cm are also generally observed and are always reported at either finite levels or at limits of detection. A complete assay
at each of the three facilities includes a measure of TRU content in the gloveboxes and HEPA filters in the glovebox exhaust.
This paper includes a description of the γ-PHA acquisitions, of the modeling, and of the calculations of nuclide content.
Because each of the remediation facilities is unique and ergonomically unfavorable to γ-ray acquisitions, we have constructed
custom detector support devices specific to each set of acquisitions. This paper includes a description and photographs of
these custom devices. The description of modeling and calculations include determination and application of container and
matrix photon energy dependent absorption factors and also determination and application of geometry factors relative to our
detector calibration geometry. The paper also includes a discussion of our measurements’ accuracy using off-line assays of
two SRNL HEPA filters. The comparison includes assay of the filters inside of 55-gallon drums using the SRNL Q2 assay system and separately using off-line assay with an acquisition configuration unique from the original in-situ acquisitions.