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This is a compilation of selected abstracts and presentations/posters from the 2010 Annual Health Physics Society (HPS) meeting held in Salt Lake City, UT and the presentations and information presented at the annual radioactive air NESHAP meeting held in conjunction with the HPS meeting. (CD-ROM)
Advanced technology and more realistic training go together to improve searches for and responses to radioactive materials.
The Office of Nuclear Incident Policy and Cooperation and the Office of Nuclear Security Deterrence and Detection at the U.S. Department of Energy, National Nuclear Security Administration provide a multi-phased approach for detection systems and measures for nuclear and other radioactive material out of regulatory control in cargo shipments. The approach is designed using best practices to deter, detect, interdict, characterize, adjudicate, and recover illicit radiation sources. Two case studies will be presented which demonstrate the multi-phased approach to nuclear security.
Certain materials used and produced in a wide range of non-nuclear industries contain enhanced activity concentrations of natural radionuclides. In particular, electricity production from coal is one of the major sources of increased human exposure to naturally occurring radioactive materials. A methodology was developed to assess the radiological impact due to natural radiation background. The developed research was applied to a specific case study, the Sines coal-fired power plant, located in the southwest coastline of Portugal. Gamma radiation measurements were carried out with two different instruments: a sodium iodide scintillation detector counter (SPP2 NF, Saphymo) and a gamma ray spectrometer with energy discrimination (Falcon 5000, Canberra). Two circular survey areas were defined within 20 km of the power plant. Forty relevant measurements points were established within the sampling area: 15 urban and 25 suburban locations. Additionally, ten more measurements points were defined, mostly at the 20-km area. The registered gamma radiation varies from 20 to 98.33 counts per seconds (c.p.s.) corresponding to an external gamma exposure rate variable between 87.70 and 431.19 nGy/h. The highest values were measured at locations near the power plant and those located in an area within the 6 and 20 km from the stacks. In situ gamma radiation measurements with energy discrimination identified natural emitting nuclides as well as their decay products (Pb-212, Pb-2142, Ra-226, Th-232, Ac-228, Th-234, Pa-234, U- 235, etc.). According to the results, an influence from the stacks emissions has been identified both qualitatively and quantitatively. The developed methodology accomplished the lack of data in what concerns to radiation rate in the vicinity of Sines coal-fired power plant and consequently the resulting exposure to the nearby population. (authors)
The release and transport characteristics of radioactive materials at a boiling accident of the high active liquid waste (HALW) in a reprocessing plant have been studied for improving experimental data of source terms of the boiling accident. In the study, a heating test and a thermogravimetry and differential thermal analysis (TG-DTA) test were conducted. In the heating test using a simulated HALW, it was found that ruthenium was mainly released into the air in the form of gas and that non-volatile elements were released into the air in the form of mist. In the TG-DTA test, the rate constants and reaction heat of thermal decomposition of ruthenium nitrosyl nitrate were obtained from TG and DTA curves. (authors)
pGamma is a code developed by the NERG group of the Technical University of Catalonia - Barcelona Tech for the analysis of gamma spectra generated by the Equipment for the Continuous Measurement and Identification of Gamma Radioactivity on Aerosols with Paper Filter developed for our group and Raditel Servies company. Nowadays the code is in the process of adaptation for the monitors of the Environmental Radiological Surveillance Network of the Local Government of Catalonia (Generalitat of Catalonia), Spain. The code is a Spectrum Analysis System, it identifies the gamma emitters on the spectrum, determines its Concentration of Activity, generates alarms depending on the Activity of the emitters and generates a report. The Spectrum Analysis System includes a library with emitters of interest, NORM and artificial. The code is being used on the three stations with the aerosol monitor of the Network (Asco and Vandellos, near both Nuclear Power Plants and Barcelona). (authors)
Two major accelerator facilities in Taiwan are about to open. The first proton treatment center equipped with a 235 MeV proton cyclotron and four gantry rooms in Linkou Chang Gung Memorial Hospital is currently under beam commissioning and clinical trials are expected to start soon. In addition, the second light source facility, named Taiwan Photon Source (TPS), in National Synchrotron Radiation Research Center is now approaching its final assembly and will be ready for beam commissioning in this year. These two facilities mark significant milestones in accelerator development and application of Taiwan. Operation of these accelerators could potentially produce intense secondary neutrons and serious material activation. Both issues are crucial and must be carefully evaluated for the purpose of radiation safety, especially in a densely populated hospital or in a crowded user experimental floor. Monte Carlo simulations with evaluated cross sections or benchmarked nuclear models are considered as the most reliable method for accelerator shielding designs or related analyses. This study predicted and compared the magnitudes of potential neutron production and induced radioactivity due to the operation of these two accelerators. For conservative consideration, the primary particle energy and full beam output were assumed to hit a thick target for maximum radiation production. In order to ensure the quality of predictions, much effort has been devoted to associated benchmark calculations, targeting those bombardment experiments involving with the same particle types and similar beam energies and targets. (authors)
The Savannah River Site stores approximately 36 million gallons of radioactive and hazardous waste that contains approximately 245 million curies. The waste is sent through various chemical processes to reduce its volume and to separate various components. The facility plans to replace formic acid (a chemical used to reduce soluble mercury) with glycolic acid. Recycle solution with glycolate may flow back to the tank farm, where the glycolate can generate hydrogen gas by thermal and radiolytic mechanisms. The current analytical method for detecting glycolate (ion chromatography) in supernatant requires a large dilution to reduce interference from the nitrate anions. Hydrogen nuclear magnetic resonance is an analytical method that requires less sample dilution. It takes advantage of the CH 2 group in glycolate. Liquid samples were spiked with four different levels of glycolate to build a calibration line, as it is recommended in the standard addition method. The detection and quantitation limits determined were 1 and 5 ppm, respectively, for 32 scans, which is well below the process limit of 10 ppm. In one test, 800 scans of a supernatant spiked with 1 ppm glycolate resulted in a -CH 2 peak with a signal-to-noise ratio of 36.
The physical principles of passive and active methods of remote-sensing monitoring of radioactive substances and toxic chemicals in accidental emissions (leaks) at nuclear fuel cycle objects are examined. The analytical possibilities of the most sensitive methods of remote-sensing monitoring, which are based on the advances made in UV and microwave range radiometry, laser IR-absorption spectroscopy, laser-induced fluorescence, and laser spark emission spectrometry, are discussed. The prospects for the development of spectral methods of remote-sensing monitoring in different ranges are analyzed.
A novel nanocomposite based on the combination of nanobentonite with nanopolyaniline and nanomanganese oxide (N-Bent–NPANI–NMn{sub 3}O{sub 4}) was synthesized and investigated to extract Co/Zn ions from water using batch technique and {sup 60}Co/{sup 65}Zn radionuclides from radioactive wastewater using column techniques. Characterization was established by High resolution-transmission electron microscopy (HR-TEM), Scanning electron microscope, X-ray diffraction (XRD), thermal gravimetric analysis (TGA) and Fourier transform infrared spectrophotometer. The image of HR-TEM indicated that the produced nanocomposite was ranged from 59.07 to 83.38 nm. Solution pH, time of reaction, amount of solid and metal concentration, all of these parameters were inspected and optimized as fundamental factors during the extraction process. The obtained results implied that the ideal conditions for extraction were pH 6 and 7, reaction time 10 and 20 min, for Co and Zn ions, respectively. The highest removal capacities were found 255.28 and 202.22 mg g{sup −1} for Co and Zn, respectively by using 5 mg nanocomposite. Excellent recoveries were obtained as 94.0–94.5% and 92.0–93.0% for removal of Co–{sup 60}Co and of Zn–{sup 65}Zn ions, respectively from the examined samples. Graphical Abstract: .
Bromine-82, Potassium-42 and Copper-64 have been successfully adopted as radioactive surrogates for outdoor large area contamination training. The goal of this project was to discover new materials that could supplement potassium bromide (KBr) and copper pellets in radiological dispersal device (RDD) training events to reduce the down time of the training fields and to broaden the toolbox of the RDD surrogate training event program at Idaho National Laboratory. Of the ten different materials investigated, sodium nitrite, gallium metal, and gallium oxide presented the greatest promise as potential materials to replace potassium bromide in RDD training events.
Here, the Fluoride-salt-cooled High-temperature Reactor (FHR) is a promising Generation IV nuclear reactor due to its improved safety design, near atmospheric working pressure, and high electric power conversion efficiency. However, control of radioactive fission gases, such as tritium, is a critical issue in FHRs due to the significantly larger production rate and potentially larger leakage rate of tritium compared to those in Light Water Reactors (LWRs). Therefore, this study proposes two mitigation strategies for tritium in FHRs: Double-Wall Heat eXchanger with a Tritium Carrier (DWHX-TC) option and Single-Wall Heat eXchanger with a Tritium Barrier (SWHX-TB) option. For the DWHX-TC option, two tube surface configurations, i.e., plain and fluted tubes, and four tritium carriers, i.e., helium, FLiBe, FLiNaK, and KF-ZrF 4 , are initially considered, among which the best tube configuration and promising tritium carrier are then selected and used for design optimization of the DWHX. For the SWHX-TB option, two cases, tritium barriers with and without cracks, are investigated in this study. In addition, the two mitigation strategies are evaluated by an in-house, one-dimensional, coupled HEat and MAss Transfer (HEMAT) code developed to predict the leakage rate of tritium in nuclear systems. Our analysis shows that (1) a double-wall configuration using fluted tubes as both the inner and outer tubes is better than other double-wall configurations investigated; (2) helium, as a tritium carrier, is superior to other candidates, including FLiBe, FLiNaK, and KF-ZrF 4 ; (3) the leakage rate of tritium (into the atmosphere) is reduced from 3400 Ci/day, taking the Advanced High-Temperature Reactor (AHTR) as a reference, to 4.0 Ci/day for an optimum design of the DWHX-TC option using helium as the tritium carrier and 1.7 Ci/day for an optimum design of the SWHX-TB option using 30-µm thick silicon carbide (SiC) as the tritium barrier (no cracks); and (4) potential cracks in tritium barriers may significantly deteriorate mass transfer performance of the SWHX-TB option. For example, a cracking rate of 0.1% leads to an increase of 694.1% in the leakage rate under the conditions investigated.
The Large Underground Xenon (LUX) experiment operates at the Sanford Underground Research Facility to detect nuclear recoils from the hypothetical Weakly Interacting Massive Particles (WIMPs) on a liquid xenon target. Liquid xenon typically contains trace amounts of the noble radioactive isotopes 85 Kr and 39 Ar that are not removed by the in situ gas purification system. The decays of these isotopes at concentrations typical of research-grade xenon would be a dominant background for a WIMP search exmperiment. To remove these impurities from the liquid xenon, a chromatographic separation system based on adsorption on activated charcoal was built. 400 kg of xenon was processed, reducing the average concentration of krypton from 130 ppb to 3.5 ppt as measured by a cold-trap assisted mass spectroscopy system. Finally, a 50 kg batch spiked to 0.001 g/g of krypton was processed twice and reduced to an upper limit of 0.2 ppt.
Microbes have been isolated previously from bentonite materials that may be used as barriers for the disposal of radioactive waste. Actively respiring microbes in such barrier materials, within a repository environment, have the potential to adversely affect waste container corrosion rates. Additionally, they could potentially alter the properties of the bentonite barrier itself. This is of significance, since the integrity of the waste container and properties of the bentonite barrier are required to fulfil defined safety functions. Furthermore to help identify the critical factors that affect microbial activity in bentonite materials, this study examines the impact of a range of parameters that could affect microbial metabolism in a geodisposal environment. Several bentonites from different sources (bentonite mined from locations in Spain and the USA, along with commercially-sourced bentonite) were subjected to increased pressure (74 MPa, 30 s), heat (90 °C, 24 h), and irradiation (1000 Gy, 24.17 Gy min –1 ), before incubation in growth media selective for sulfate-reducing bacteria (SRB) or iron-reducing bacteria (IRB).
Repositories for high level radioactive waste hosted in argillite or granite with bentonite backfill are the two disposal concepts that been extensively studied. A general perception is that granite has very low sorption capacity and therefore is disadvantageous in retarding the migration of radionuclides in comparison with argillite. However, when there is bentonite backfill, does the sorption capacity of granite really matter for the migration of radionuclides? How different host rocks would affect the migration of U(VI) through the bentonite backfill of an engineered barrier system (EBS), and which properties of the host rocks cause those differences in migration, given that migration of radionuclides is a critical measure in assessing the performance of a repository? Here we present two coupled thermal, hydrological, and chemical models for the transport of U(VI), each with an identical setup except for the type of host rock (argillite vs granite rock). Comparisons between our models show that the different host rocks exert their influence on migration of U(VI) via regulating the chemical conditions in the bentonite, which affect the concentration of U(VI) at the source and the adsorption of U(VI) in the bentonite. The key chemical conditions are pH, Ca + , and HCO 3 - concentration in pore-water, which are strongly affected by soluble carbonate minerals. Our models also show the occurrence of illitization (dissolution of smectite and precipitation of illite) in the bentonite, which affects the migration of U(VI) through changing pH and the quantity of adsorbents. Although generalization of current model results for granite and argillite should be done carefully, the simulations highlight the importance of pore-water chemistry in host rock, as well as bentonite-host rock interactions, when assessing the performance of a repository. In comparison to properties such as thermal conductivity, permeability, and adsorption capacity, properties that have been studied intensively, the pore-water chemistry of host rocks deserve equal or even more attention because of its strong influence on radionuclide migration.
Corrosion-resistant welded alloys are frequently used as a leak-tight boundary in critical applications that require confinement of hazardous and/or radioactive substances, including an increasing population of spent nuclear fuel (SNF) canisters. The behavior of residual stresses generated as a result of irregular elastic–plastic deformation during processes such as welding is one of today's key issues to a full understanding of the aging mechanisms that may compromise the confinement boundary. Whether such processes and any subsequent weld repairs, not subjected to post-weld heat treatment, would negatively affect the initial material by introducing through-thickness tensile stresses remains an open question. Here we report the first residual stress measurements using neutron diffraction on the welded joints of a SNF canister. We found significant tensile residual stresses in the as welded sample, indicating that initiation and through-thickness growth of cracks may be possible. Following repair, we observed a stress redistribution and introduction of beneficial compressive stresses. We anticipate our results will improve understanding of confinement susceptibility to aging and guide improvements in repair techniques.
A detection technique for high-efficiency measurements of fusion reactions with low-intensity radioactive beams was developed. The technique is based on a Multi-Sampling Ionization Chamber (MUSIC) operating as an active target and detection system, where the ionization gas acts as both target and counting gas. In this way, we can sample an excitation function in an energy range determined by the gas pressure, without changing the beam energy. The detector provides internal normalization to the incident beam and drastically reduces the measuring time. In the first experiment we tested the performance of the technique by measuring the 10,13,15 C+ 12 C fusion reactions at energies around the Coulomb barrier.
Here, we report results from a systematic measurement campaign conducted to identify low radioactivity materials for the construction of the EXO-200 double beta decay experiment. Partial results from this campaign have already been reported in a 2008 paper by the EXO collaboration. Here we release the remaining data, collected since 2007, to the public. The data reported were obtained using a variety of analytic techniques. The measurement sensitivities are among the best in the field. We concluded the construction of the EXO-200 detector and Phase-I data was taken from 2011 to 2014. The detector’s extremely low background implicitly verifies the measurements and the analysis assumptions made during construction and reported in this paper.