Validating a Methodology That Associates Minimum Detectable Activity with Detector Velocity.

ABSTRACT: Mobile radiation detection systems are used widely in remediation and nuclear security. However, their detection efficiency and thus their minimum detectable activity is not completely understood. It is recognized that the detector's velocity will affect its detection efficiency. At lower speeds, detection efficiency will be higher than a detector moving at higher speeds. The relationship describing how speed and efficiency are related was only recently quantified using a modified four-parameter logistic function (M4PL). The current research verifies the M4PL function in a controlled laboratory setting using a 5.08 cm × 5.08 cm sodium iodide detector at speeds between 20-120 cm s-1. As expected, the M4PL function indicates a detection system's highest efficiency at low speeds and its lowest efficiency at higher speed. In between is a transition region of decreasing detector efficiency. This decrease is gradual within initial speeds but quickly steepens and then approaches a minimum at higher detector speeds. This general shape was observed in the experimental data and validated the M4PL function as a predictive tool. To conclude this research and to demonstrate the function's usefulness, a relationship between speed and minimum detectable activity (MDA) was developed. Using this function and the methods described in this research, planners can now optimize surveys by controlling velocity to maintain MDA. It is also possible to use this technique to accelerate surveys while having the ability to predict the reduction in MDA. This foundational relationship between detector speed and detection efficiency has the potential to improve detector performance in various applications for both the academic and operational fields.

Validation of a Dose Assessment Method to be Used in 18F FDG Loose Contamination Exercises.

ABSTRACT: Radiological emergency response may require responders to operate in contaminated environments. To provide more realistic training to these individuals, it has been proposed to disperse low amounts of short-lived radioactive material in simulated emergency scenarios. To demonstrate the applicability and safety of such activities, a limited exercise was conducted where 18F was sprayed in a small area and survey activities were executed. A pre-job external radiation exposure dose assessment was performed in preparation for this training. The research presented here compares participant external recorded doses to assessment results in order to validate the dose estimates. Two individuals were used during the dispersion, search, and survey activities. First, a radiation worker mixed 200 MBq Fludeoxyglucose 18F with 470 mL H2O in a weed sprayer and distributed it over a 3 m × 3 m area. After evaporation, an exercise participant performed search and survey activities in the area. Actual whole-body doses measured with optically stimulated luminescence dosimeters were 10 ± 1 µSv for both personnel. Whole-body digital dosimeters read 4.3 ± 0.2 µSv and 3.3 ± 0.5 µSv for the radiation worker and exercise participant, respectively. Actual extremity doses were below the dosimeters' minimum detectable limits for the radiation worker (thermoluminescence dosimeter) and exercise participant (optically stimulated luminescence dosimeter). The dose assessment-predicted whole-body doses were 2.8 ± 0.4 µSv and 3.2 ± 0.1 µSv for the radiation worker and exercise participant, respectively. The estimated dose to the radiation worker's hand was 21.8 ± 3.8 µSv, and the estimated dose to the exercise participant's knee was 13.4 ± 0.6 µSv. The study provided substantial evidence for the validity of the dose assessment method, supporting its use for a larger training exercise.

Preliminary Dose Assessment for Emergency Response Exercise Using Unsealed Radioactive Contamination.

A preliminary dose assessment for an emergency response exercise using unsealed radioactive sources was performed based on conservative calculation methods. The assessment was broken into four parts: activation, distribution, exercise participation, and post-exercise monitoring. The computer code MicroShield was used to determine external exposure from the source during and after distribution. Internal exposure via inhalation and ingestion was estimated by assuming fractional intakes of activity and converting to dose using annual limits on intake and dose coefficients. It was determined from the dose assessment that a radionuclide-dependent range of 37 MBq to 1.5 GBq can be used to achieve detectable dose rates during the exercise without exceeding assumed administrative dose limits. Of the identified radionuclides, Tc results in the lowest dose and is recommended from a radiological safety standpoint. However, the choice of which radionuclide and what activity to use for an exercise should be made based on budget and the logistics of the actual exercise.

Radionuclide Selection for Emergency Response Exercise at Disaster City® Using Unsealed Radioactive Contamination.

The Department of Nuclear Engineering at Texas A&M University currently supports exercises at Disaster City, a mock community used for emergency response training that features full-scale, collapsible structures designed to simulate various levels of disaster and wreckage. Emergency response exercises can be enhanced by using unsealed radioactive sources to simulate a more realistic response environment following an incident involving the dispersion of radioactive material. Limited exercises are performed worldwide using unsealed radioactive sources, and most of that information is not publicly available. This research compiles the publicly available information along with additional information acquired through discussion with experts and presents the process for selection of a short-lived radionuclide for use at Disaster City. The historically-used radionuclides were F, Tc, Br, and La. These radionuclides were considered for the Disaster City exercise, as well as other short-lived radionuclides commonly used or capable of being produced at Texas A&M. The selection process described in this paper identified seven radionuclides that could be used in an unsealed contamination exercise at Disaster City. Radiopharmaceuticals Tc and F are suitable and available for purchase from nearby vendors. In addition, the Texas A&M Nuclear Science Center TRIGA reactor could be used to produce Na, Mn, Cu, Br, and La via thermal neutron activation.

Calculation of Canine Dose Rate Conversion Factors for Photons and Electrons.

Urban search and rescue (USAR) dogs are valuable members of their teams and play key roles in performing successful missions. A pair of dogs can do the work of dozens of people, the dogs are able to quickly sniff around collapsed structures and zip through constricted hallways with far greater accuracy than their plodding human counterparts. While in contaminated areas, their human counterparts are afforded the benefit of personal protective equipment (PPE) to keep exposures to chemical, biological and radiological substances to a minimum; USAR dogs, on the other hand, are not. In an effort to allow USAR dogs to be used to their full potential, PPE is often not worn as it inhibits their ability to move in and around obstacles to use their strong senses of smell and hearing. In addition, these animals may snag or be snagged on debris or structures, which may require rescue of the animal. In a collaborative effort between Texas A&M University's Department of Nuclear Engineering and the College of Veterinary Medicine, researchers are attempting to estimate the extent of the radiation doses received by these valuable team members during missions where radioactive contamination is present. Currently there are no dose rate conversion factors for USAR dogs, and those that are available are calculated at a height of 1 m. To address this issue, a more suitable height of 40 cm was chosen, and dose conversion factors were calculated for monoenergetic photon sources ranging from 15 keV to 15 MeV and for monoenergetic electron sources ranging from 10 keV to 10 MeV. The radioactivity is assumed to be uniformly distributed on the surface of the ground. Forty centimeters was chosen as the height of interest for the three breeds FEMA prefers as USAR dogs. These dose conversion factors will permit dose estimates to be made, allowing these animals to do their jobs successfully while keeping their radiation doses as low as possible.

Modeling Study of a Proposed Field Calibration Source Using K-40 and High-Z Targets for Sodium Iodide Detectors.

Calibration sources based on the primordial isotope potassium-40 (K) have reduced controls on the source's activity due to its terrestrial ubiquity and very low specific activity. Potassium-40's beta emissions and 1,460.8 keV gamma ray can be used to induce K-shell fluorescence x rays in high-Z metals between 60 and 80 keV. A gamma ray calibration source that uses potassium chloride salt and a high-Z metal to create a two-point calibration for a sodium iodide field gamma spectroscopy instrument is thus proposed. The calibration source was designed in collaboration with the Sandia National Laboratory using the Monte Carlo N-Particle eXtended (MCNPX) transport code. Two methods of x-ray production were explored. First, a thin high-Z layer (HZL) was interposed between the detector and the potassium chloride-urethane source matrix. Second, bismuth metal powder was homogeneously mixed with a urethane binding agent to form a potassium chloride-bismuth matrix (KBM). The bismuth-based source was selected as the development model because it is inexpensive, nontoxic, and outperforms the high-Z layer method in simulation. Based on the MCNPX studies, sealing a mixture of bismuth powder and potassium chloride into a thin plastic case could provide a light, inexpensive field calibration source.

Developing a Methodology for Determination of Elemental Composition of Shielding Materials.

Radiation transport simulation models can provide estimations of radiation effects such as detector response and detection capabilities. The objective of this research was to develop a methodology for quick, efficient, and effective determination of the composition of shielding materials to be used in radiation transport models. A C++ code, MatFit, was developed that used the concept of densitometry and the iterative method developed for the Spectrum Analysis by Neutron Detectors II (SAND II) computer program to estimate the elemental composition of shielding materials. These results were compared to previous neutron activation analysis (NAA) on the same samples. It was determined that densitometry provided an elemental approximation that yielded an attenuation rate within 10% of that found through NAA but requires much fewer resources, as well as less time. From this research, it is recommended that the developed method and C++ program be used when constructing models for detector response.

Signal Processing and Its Effect on Scanning Efficiencies for a Field Instrument for Detecting Low-energy Radiation.

Signal processing within a radiation detector affects detection efficiency. Currently, organizations such as private industry, the U.S. Navy, Army, and Air Force are coupling some detector systems with data collection devices to survey large land areas for radioactive contamination. As detector technology has advanced and analog data collection has turned to digital, signal processing is becoming prevalent in some instruments. Using a NIST traceable (241)Am source, detection efficiency for a field instrument for detecting low-energy radiation (FIDLER) was examined for both a static and scanning mode. Experimental results were compared to Monte Carlo-generated efficiencies. Stationary data compared nicely to the theoretical results. Conversely, scanning detection efficiencies were considerably different from their theoretical counterparts. As speed increased, differences in detection efficiency approached two orders of magnitude. To account for these differences, a quasi time-dependent Monte Carlo simulation was created mimicking the signal processing undertaken by the FIDLER detection system. By including signal processing, experimental results fell within the bounds of the Monte Carlo-generated efficiencies, thus demonstrating the negative effects of such processing on detection efficiencies.

List mode with the ORTEC digiBASE-E.

The ORTEC digiBASE-E (ORTEC, 801 S. Illinos Ave., Oak Ridge TN 37831) is a compact photomultiplier tube endcap designed to handle all of the necessary power and signal processing requirements of a scintillation gamma-ray detector. The list mode feature of this device was used by a custom software package (CraneWow, Department of Nuclear Engineering, Texas A&M University, College Station, TX 77843) to gather data during seaport operations unloading cargo containers. A number of difficulties were encountered in creating the software and are catalogued here to aid future users of the device.

Predicting instrument detection efficiency when scanning point and small area radiation sources.

Accurate quantification of radionuclides detected during a scanning survey relies on an appropriately determined scan efficiency calibration factor (SECF). Traditionally, instrument efficiency is determined with a stationary instrument and a fixed source geometry. However, as is often the case, the instrument is used in a scanning mode where the source to instrument geometry is dynamic during the observation interval. Procedures were developed to determine the SECF for a point source ("hot particle") and a 10 x 10 cm source passing under the centerline of a 12.7 x 7.62 cm NaI(Tl) detector. The procedures were first tested to determine the SECF from a series of static point source measurements using Monte Carlo N-Particle code. These point static efficiency values were then used to predict the SECF for scan speeds ranging from 10 cm s(-1) to 80 cm s(-1) with a simulated instrument set to collect integrated counts for 1 s. The Monte Carlo N-Particle code was then used to directly determine the SECF by simulating a scan of a point source and 10 x 10 cm area source for scan speeds ranging from 10 cm s(-1) to 80 cm s(-1). Comparison with Monte Carlo N-Particle scan simulation showed the accuracy of the SECF prediction procedures to be within +/-5% for both point and area sources. Experimental results further showed the procedures developed to predict the actual SECF for a point and 10 x 10 cm source to be accurate to within +/-10%. Besides the obvious application to determine an SECF for a given scan speed, this method can be used to determine the maximum detector or source velocity for a desired minimum detectable activity. These procedures are effective and can likely be extended to determine an instrument specific SECF for a range of source sizes, scan speeds, and instrument observation intervals.