Quantification of a Facility Radiological Security Risk Index With a Graphical User Interface Tool.

ABSTRACT: Healthcare facilities around the world routinely use radioactive sources to diagnose and treat illness. To effectively manage the security of radioactive sources, these facilities need to understand the risk, which is comprised of threat, vulnerability, and consequences. The threat component of risk requires knowledge of potential adversaries and understanding their capabilities and intentions. To help articulate the multiple layers of threat and support better informed decisions, the research developed a risk-based methodology to evaluate radiological security at the facility level. The methodology is applied to a radiological dispersal device (RDD) incident from three radionuclides of concern: 137Cs, 60Co, and 192Ir. The results of the research have led to the creation of a potential facility risk index (PFRI). The PFRI is mathematically represented as the exponential product of the maximum expected utility among the threat groups, the sum of geographic vulnerability and cultural vulnerability, and net consequences. The PFRI is a novel risk index that quantifies the facility risk on a scale of 1 to 10, 1 being "very low risk" and 10 being "very high risk." A MATLAB-based graphical user interface (GUI) tool was also developed to enable the radiological facility (i.e., healthcare facility) staff to conduct self-assessments and manage their most valuable assets. The PFRI methodology is a useful starting point for any healthcare facility risk assessment and is a valuable input for decision makers considering potential investments in security upgrades.

Nuclear and Radiological Source Security Culture Assessment of Radioactive Material Users at a University.

Securing radioactive sources has become increasingly important given the rising threat of radiological terrorism. While radiation safety has long been established in most applicable industries, the importance of nuclear and radiological source security has lagged behind in nonnuclear material specific industries, such as academic institutions and medical facilities. To evaluate the attitudes and behaviors regarding nuclear security culture, an assessment of nuclear and radiological material practices was developed and conducted on 73 radioactive material users at a university. The survey portion of the assessment was comprised of a series of questions segregated into four categories: policy, enforcement, leadership, and behavior. Nuclear security awareness questions formed a subset of the questionnaire. Users were classified by their radioactive material experience and work classification: student, faculty, or other staff. Of the users surveyed, 9% were also interviewed face-to-face to further expand on their views of nuclear security culture. Results of the assessment showed that students from the work classification group as well as the cohort of radioactive material users with 2-5 y of experience possessed a greater degree of awareness towards nuclear security compared to faculty and other more experienced radioactive material users. Relative to students and faculty, other staff from the work classification group faced some difficulty judging the enforcement of policies, adequacy of inspection, and job performance review related to nuclear security. The response from all three groups emphasized the need to enhance threat-response preparedness and greater communication among stakeholders.

Biokinetics of Plutonium in Nonhuman Primates.

A major source of data on metabolism, excretion and retention of plutonium comes from experimental animal studies. Although old world monkeys are one of the closest living relatives to humans, certain physiological differences do exist between these nonhuman primates and humans. The objective of this paper was to describe the metabolism of plutonium in nonhuman primates using the bioassay and retention data obtained from macaque monkeys injected with plutonium citrate. A biokinetic model for nonhuman primates was developed by adapting the basic model structure and adapting the transfer rates described for metabolism of plutonium in adult humans. Significant changes to the parameters were necessary to explain the shorter retention of plutonium in liver and skeleton of the nonhuman primates, differences in liver to bone partitioning ratio, and significantly higher excretion of plutonium in feces compared to that in humans.

(47)Ca production for (47)Ca/(47)Sc generator system using electron linacs.

In this work we have studied the feasibility of photonuclear production of (47)Ca from (48)Ca for (47)Ca/(47)Sc generators. Photon flux distribution for electron beams of different energies incident on a tungsten converter was calculated using the MCNPX radiation transport code. The (47)Ca production rate dependence on electron beam energy was found and (47)Ca/(47)Sc yields were estimated for a 40MeV electron beam. It was shown that irradiating enriched targets with a 40MeV, 1mA beam will result in tens of MBq g(-1) (few mCi g(-1)) activity of (47)Sc. The results of the simulations were benchmarked by irradiating 22.5g of CaCl2 powder with a 39MeV electron beam incident on a tungsten converter. Measured (47)Ca/(47)Sc activities were found to be in very good agreement with the predictions.

Particle size characterization of aerosols generated during surface contaminated concrete demolition.

The purpose of this study was to measure mass and activity distributions from the aerosols generated during the demolition of surface contaminated concrete. Air samples were collected using a cascade impactor during the mechanical hammering and dismantlement of radiologically contaminated high level waste process vaults from which the piping and components had been previously removed. The experimentally determined distribution parameters were compared with the 5.0-µm particle size referenced in the U.S. Department of Energy (U.S. DOE) regulations pertaining to internal deposition (10CFR835 and ICRP66). Mass distribution parameters were compared with their activity distribution counterparts. The Mass Median Aerodynamic Diameter (MMAD) was determined to be 4.2 µm with a Mass Geometric Standard Deviation (GSDM) of 2.3 µm, and the Co Activity Median Aerodynamic Diameter (AMAD) was determined to be 3.9 µm with an Activity Geometric Standard Deviation (GSDA) of 2.3 µm. These results are consistent with the ICRP66 5.0-µm reference particle size and the Derived Air Concentration (DAC) values referenced in 10CFR835 and utilized throughout the U.S. DOE complex.

Monte Carlo simulation of in vivo measurement of the most suitable knee position for the optimal measurement of activity.

A new computational model has been developed using the Monte Carlo (MC) technique to simulate in vivo measurements with the objective of understanding the most precise measurement location with respect to quantifying the activity of Am in the bones. To benchmark the model, in vivo measurements were performed on the U.S. Transuranium and Uranium Registries (USTUR) case 0846 leg. Front and lateral measurements of the knee of the USTUR case 0846 leg in a bent position and the same measurements with the leg in a straight position using a HP(Ge) detector were completed. Experimental results concluded that the front measurement of the knee in a bent leg position gave the highest count rate, which is an indication of optimal detection efficiency. Therefore, this geometry and knee-detector position were considered as the most appropriate position for knee monitoring. A computational model using MCNPX version 2.6.0 was used to simulate the experimental measurements by using a leg voxel phantom. The mean value and standard deviation (SD) of peak efficiency due to an isotropic 59.5-keV photon from Am were calculated in four different counting geometries. An extra sum of squares F-test was performed on the mean values of the simulated detection efficiencies. The p-value obtained from this statistical test indicates that the differences among the mean values for different counting geometries were significant. These results suggest that the front measurement of a knee in a bent leg position is the optimal counting geometry for in vivo measurement of Am deposited in the bones. The computational model was validated through comparison of the measured and simulated detection efficiencies. It was observed that there is no difference at the 0.1 significant levels between the simulated and measured detection efficiencies in assessment of Am within the bones.

Radiological effluents released by U.S. commercial nuclear power plants from 1995-2005.

Commercial nuclear power plants release gaseous and liquid radiological effluents into the environment as by-products of electrical generation. In the U.S. these releases are monitored by the Nuclear Regulatory Commission (U.S. NRC) and Environmental Protection Agency (U.S. EPA). Traditionally these releases have always been well below the regulatory limits. However, the tracking and analysis of nuclear power radiological effluents was stopped in 1994 by several government agencies. The purpose of this study was to compile the entire U.S. industry effluent data, identify trends, and calculate average population dose commitments since that time. Data were taken from radioactive material release reports submitted by each nuclear power plant. Industry trends were identified using the Mann-Kendall non-parametric test. Total collective effective and population doses were estimated using UNSCEAR and U.S. NRC methodologies. Overall, industry releases have been level over the study time period. Public doses continue to be well below 1% of the regulatory limits.

Tritium recapture behavior at a nuclear power reactor due to airborne releases.

This paper describes the initiatives taken by Cook Nuclear Plant to study the on-site behavior of recaptured tritium released in its airborne effluents. Recapture is the process where a released radioactive effluent, in this case tritium, is brought back on-site through some mechanism. Precipitation, shifts in wind direction, or anthropogenic structures that restrict or alter effluent movement can all lead to recapture. The investigation was started after tritium was detected in the north storm drain outfall. Recent inadvertent tritium releases by several other nuclear power plants, many of which entered the groundwater, have led to increased surveillance and scrutiny by regulatory authorities and the general public. To determine the source of tritium in the outfall, an on-site surface water, well water, rainwater and air-conditioning condensate monitoring program was begun. Washout coefficients were also determined to compare with results reported by other nuclear power plants. Program monitoring revealed detectable tritium concentrations in several precipitation sample locations downwind of the two monitored containment building release vents. Tritium was found in higher concentrations in air-conditioning condensate, with a mean value of 528 Bq L(-1) (14,300 pCi L(-1)). The condensate, and to a lesser extent rainwater, were contributing to the tritium found in the north storm drain outfall. Maximum concentration values for each sample type were used to estimate the most conservative dose. A maximum dose of 1.1 x 10(-10) mSv (1.1 x 10(-8) mrem) total body was calculated to determine the health impact of the tritium detected.