[en] The Analytical Development Section of Savannah River Technology Center (SRTC) was requested to determine the holdup of enriched uranium in a facility as part of an overall deactivation project of the facility. The 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 and Accountability, and to meet criticality safety controls. This report covers calibration of the detectors in order to support holdup measurements in the 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 U-235, Np-237, and Am-24 1 that were observed in these components. The detector system was run by an EG and G system that contains the high voltage power supply and signal processing electronics. A personal computer with Gamma-Vision software was used to control and provide space to store and manipulate multiple channel spectra. The measured Np-237 and Am-241 contents were especially important in these components because their presence is unusual and unexpected in the facility. 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 a solid waste vault. This report presents determination of the calibration constants from first principles for determination of Am-241 and Np-237 using this detection system and compares the values obtained for Np-237 with the calibration factors obtained with a subsequent measurement using a point source of radioactive equilibrium Np-237/Pa-233

[en] The sensitivity of environmental sampling and analysis for the estimated atmospheric concentrations of radionuclides in effluents from clandestine nuclear facilities may require many sampling sites and exorbitant costs to monitor such large areas

[en] The use of HPGe detectors in counting situations where the sample is not easily reproduced has increased the use of models to determine the counting efficiency for the specific geometry. The accuracy of these simulations of the germanium detector response relies on detailed knowledge of the performance of the detector. Several different types of detectors were measured at different energies using a pencil beam of gamma-rays. These measurements showed that the dead layer was not uniform from detector to detector. This and the construction details were used to calculate the efficiency for several detectors

[en] A radiochemical isotope dilution mass spectrometry method has been developed to determine the age of uranium materials. The amount of 230Th activity, the first progeny of 234U, that had grown into a small uranium metal sample was used to determine the elapsed time since the material was last radiochemically purified. To preserve the sample, only a small amount of oxidized uranium was removed from the surface of the sample and dissolved. Aliquots of the dissolved sample were spiked with 233U tracer and radiochemically purified by anion-exchange chromatography. The 234U isotopic concentration was then determined by thermal ionization mass spectrometry. Additional aliquots of the sample were spiked with 229Th tracer, and the thorium was purified using two sequential anion-exchange chromatography separations. The isotopic concentrations of 230Th and 232Th were determined by TIMS. The lack of any 232Th confirmed the assumption that all thorium was removed from the uranium sample at the time of purification. The 230Th and 234U mass concentrations were converted to activities and the 230Th/234U ratio for the sample was calculated. The experimental 230Th/234U ratio showed the uranium in this sample was radiochemically purified in about 1945. Isotope dilution thermal ionization mass spectrometry has sufficient sensitivity to determine the age of 100 samples of uranium. This method could certainly be employed as a nuclear forensic method to determine the age of small quantities of uranium metal or salts. Accurate determination of the ultra-trace 230Th radiochemically separated from the uranium is possible due to the use of 229Th as an isotope dilution tracer. The precision in the experimental age of the uranium could be improved by making additional replicate measurements of the 230Th/234U isotopic ratio or using a larger initial sample

[en] Batch equilibrium measurements were conducted with a granular Resorcinol-Formaldehyde (RF) resin to determine the distribution coefficients (Kds) for cesium. In the tests, Hanford Site actual waste sample containing radioactive cesium and a pretreated waste sample that was spiked with non-radioactive cesium were used. Initial concentrations of non-radioactive cesium in the waste sample were varied to generate an equilibrium isotherm for cesium. Two additional tests were conducted using a liquid to solid phase ratio of 10 and a contact time of 120 hours. The measured distribution coefficient (Kd) for radioactive cesium (137Cs) was 948 mL/g; the Kd for non-radioactive cesium (133Cs) was 1039 mL/g. The Kd for non-radioactive cesium decreased from 1039 to 691 mL/g as the initial cesium concentration increased. Very little change of the Kd was observed at initial cesium concentrations above 64 mg/mL. The maximum sorption capacity for cesium on granular RF resin was 1.17 mmole/g dry resin. T his value was calculated from the fit of the equilibrium isotherm data to the Dubinin-Radushkevich equation. Previously, a total capacity of 2.84 mmole/g was calculated by Bibler and Wallace for air-dried RF resin

[en] Matrix-assisted laser desorption/ionization coupled with time-of-flight mass spectrometry (MALDI/TOF-MS) was used for the analysis of low-molecular phosphate compounds found in Hanford tank wastes. The mass spectra of these compounds indicate protonated peaks as well as sodium adducts. Analytical methods presently utilized for the analysis of the phosphate-related organics are both time consuming and labor intensive. A promising alternative is MALDI/TOFMS. The MALDI process produces both positive and negative ions directly and very little sample is required. In addition,there is limited sample preparation and minimal hazardous waste production

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[en] The United States, the Department of Energy (DOE) and its National Laboratories, including the Pacific Northwest National Laboratory (PNNL), are facing a serious attrition of nuclear scientists and engineers and their capabilities through the effects of aging staff. Within the DOE laboratories, 75% of nuclear personnel will be eligible to retire by 2010. It is expected that there will be a significant loss of senior nuclear science and technology staff at PNNL within five years. PNNL's nuclear legacy is firmly rooted in the DOE Hanford site, the World War II Manhattan Project, and subsequent programs. Historically, PNNL was a laboratory were 70% of its activities were nuclear/radiological, and now just under 50% of its current business science and technology are nuclear and radiologically oriented. Programs in the areas of Nuclear Legacies, Global Security, Nonproliferation, Homeland Security and National Defense, Radiobiology and Nuclear Energy still involve more than 1,000 of the 3,800 current laboratory staff, and these include more than 420 staff who are certified as nuclear/radiological scientists and engineers. This paper presents the current challenges faced by PNNL that require an emerging strategy to solve the nuclear staffing issues through the maintenance and replenishment of the human nuclear capital needed to support PNNL nuclear science and technology programs

[en] A fully automated analysis procedure and instrument for the measurement of total 99Tc in aged nuclear waste has been developed. The overall analysis approach is based on a fully automated wet radiochemical analysis method. Microwave-assisted sample oxidation is used prior to a chemical separation step in order to oxidize all of the non-pertechnetate species to pertechnetate. Separation of the pertechnetate from interfering radioactive and stable matrix species is carried out using anion exchange column. The separated 99Tc is quantified using a flow-through solid cell scintillation detector. The instrument is capable of an analysis time of <13 minute per sample with a detection limit of 2000 dpm/mL. Nuclear waste samples from the Hanford site with a high content of non-pertechnetate species were successfully analyzed using this method