Design and Characterization of an Extremely-Sensitive, Large-Volume Gamma-Ray Spectrometer for Environmental Samples.

A large volume gamma spectrometer was designed and constructed to analyze foodstuffs and environmental samples having low radionuclide concentrations. This system uses eight 11-cm × 42.5-cm × 5.5-cm NaI(Tl) detectors, chosen due to their relatively high sensitivity and availability and arranged in an octagonal configuration. The sensitive volume of the system is ~28 cm in diameter and ~42 cm deep. Shielding consists of an 86-cm × 86-cm square, 64-cm-tall lead brick enclosure with 18-cm-thick lead walls lined by 0.3-cm-thick copper plates. An aluminum top was machined to suspend the detectors within this shield. The shielding reduces background counts by 72% at 100 keV and 42% at 1,000 keV. The positional variability in sensitivity of the well was determined by both simulation and experiment. A 2.1-L volume of nearly uniform sensitivity, varying less than 10%, exists in the well's center. Energy resolutions of 14.6% and 7.8% were measured for Am and Cs, respectively. Energy resolution shows a 0.2% variation for both Am and Cs as a function of position within all regions of the well's central sensitive volume. Dead time was also determined to be less than 35% for all sources measured in the system, the largest of which had an activity of 1,760 kBq. Simulated results for various source geometries show higher counts for smaller samples, especially at lower energies due to less attenuation of low energy photons. Minimum detectable activities were determined for all source energies used, less than 5.1 Bq kg for reasonable background and sample counting times.

The Effects of Radiation and Emitted Light Transport on the Positional Response of 11 cm × 42.5 cm × 5.5 cm NaI(Tl) Detectors.

Experiments were performed with 30 11 cm × 42.5 cm × 5.5 cm NaI(Tl) detectors to better understand their positional response. Spectra were collected using 0.02 to 0.15 MBq point sources of Am, Cs, Co, and Ba positioned on lines parallel and perpendicular to the long axis of the crystal along both the narrow and wide detector faces as well as at different distances from them. A greater density of positions was sampled at the ends of the detector, and repeated measurements were made to examine potential gain drifts during the experiment. Spectroscopic peak counts, spectroscopic pulse heights, and net counts were analyzed. Empirical equations were fit to the aforementioned data for each specific source energy as a function of source position. In addition, a Monte Carlo radiation transport code was used to simulate the expected positionally variable response based solely upon radiation absorption. The simulated radiation transport efficiency functions were compared to the experimental data. The effects of the geometric radiation efficiency, the attenuation and scattering of emitted light within the scintillation crystal, and combined effects such as nonuniformity of the photomultiplier tube, photocathode response, and crystal irregularities were then distinguished. Functions describing each effect were derived. The results suggest potential new corrections to data obtained with large scintillation detectors as well as a novel approach to partial positional gamma-ray detection with minimal collimation, given that the energy resolution is within reason for particular photopeaks.

Radon dispersion modeling and dose assessment for uranium mine ventilation shaft exhausts under neutral atmospheric stability.

In the present study, the roles of atmospheric wind profiles in the neutral atmosphere and surface roughness parameters in a complex terrain were examined to determine their impacts on radon ((222)Rn) dispersion from an actual uranium mine ventilation shaft. Simulations were completed on (222)Rn dispersion extending from the shaft to a vulnerable distance, near the location of an occupied farmhouse. The eight dispersion scenarios for the ventilation shaft source included four downwind velocities (0.5, 1.0, 2.0 and 4.0 m s(-1)) and two underlying surface roughness characteristics (0.1 m and 1.0 m). (222)Rn distributions and elevated pollution regions were identified. Effective dose estimation methods involving a historical weighting of wind speeds in the direction of interest coupled to the complex dispersion model were proposed. Using this approach, the radiation effects on the residents assumed to be outside at the location of the farm house 250 m downwind from the ventilation shaft outlet were computed. The maximum effective dose rate calculated for the residents at the outside of the farm house was 2.2 mSv y(-1), which is less than the low limit action level of 3-10 mSv y(-1) recommended by the International Commission on Radiological Protection (ICRP) occupational exposure action level for radon.

Comparison of I-131 radioimmunotherapy tumor dosimetry: unit density sphere model versus patient-specific Monte Carlo calculations.

High computational requirements restrict the use of Monte Carlo algorithms for dose estimation in a clinical setting, despite the fact that they are considered more accurate than traditional methods. The goal of this study was to compare mean tumor absorbed dose estimates using the unit density sphere model incorporated in OLINDA with previously reported dose estimates from Monte Carlo simulations using the dose planning method (DPMMC) particle transport algorithm. The dataset (57 tumors, 19 lymphoma patients who underwent SPECT/CT imaging during I-131 radioimmunotherapy) included tumors of varying size, shape, and contrast. OLINDA calculations were first carried out using the baseline tumor volume and residence time from SPECT/CT imaging during 6 days post-tracer and 8 days post-therapy. Next, the OLINDA calculation was split over multiple time periods and summed to get the total dose, which accounted for the changes in tumor size. Results from the second calculation were compared with results determined by coupling SPECT/CT images with DPM Monte Carlo algorithms. Results from the OLINDA calculation accounting for changes in tumor size were almost always higher (median 22%, range -1%-68%) than the results from OLINDA using the baseline tumor volume because of tumor shrinkage. There was good agreement (median -5%, range -13%-2%) between the OLINDA results and the self-dose component from Monte Carlo calculations, indicating that tumor shape effects are a minor source of error when using the sphere model. However, because the sphere model ignores cross-irradiation, the OLINDA calculation significantly underestimated (median 14%, range 2%-31%) the total tumor absorbed dose compared with Monte Carlo. These results show that when the quantity of interest is the mean tumor absorbed dose, the unit density sphere model is a practical alternative to Monte Carlo for some applications. For applications requiring higher accuracy, computer-intensive Monte Carlo calculation is needed.

A method for the quantitative gamma spectroscopic analysis of an unusually shaped unknown source.

An unmarked cylindrical device, identified as a ceramic high voltage capacitor, needed its radioactivity assessed so that proper disposal and shipping requirements could be met. Using a high purity germanium detector, naturally occurring 232Th was identified as the source of radioactivity. A series of point source measurements was made along the length of the item's axis using 60Co, having a gamma ray of nearly the same energy as one of the primary 232Th progeny photopeaks. These measurements were then numerically integrated to determine the response of the detector to a line source. A correction for the self shielding of the item was estimated using Monte Carlo simulations. The item was found to contain approximately 1.85 x 10(5) Bq of uniformly distributed 232Th. The overall method has application to any unusually shaped source, with point source measurements performed using an appropriate radionuclide used to establish an overall sensitivity of the detector, including its dead layer, to the radioactivity in a simple geometric representation of the object. An estimation of self shielding from Monte Carlo is then applied to that result.

Simulation, design, and construction of a 137Cs irradiation facility.

Regulatory entities require that for any radiation facility the surrounding areas must be restricted unless the dose equivalent is less than 0.02 mSv in any one hour. Two Monte Carlo radiation transport simulation codes, MCNP5 and Mercurad, were used to design a facility to shield a 3.48 x 10(5) MBq 137Cs irradiator that meets these requirements. Simulations showed that the dose equivalent rates were below the legal limit for unrestricted access and the facility was constructed using available concrete block and student labor to minimize costs. To verify the accuracy of the Monte Carlo radiation transport codes, an ion chamber was used to characterize the facility. Ion chamber measurements in the actual, as-built irradiation facility showed that the Monte Carlo codes, MCNP5 and Mercurad, agreed by a factor of better than 6% and better than 11%, respectively.

Design and simulation of a neutron facility.

State and other regulatory entities require that for any facility housing a particle accelerator the surrounding areas must be restricted to public access unless the dose equivalent rate is less than 0.02 mSv h at 5 cm from any accessible wall surrounding the facility under conditions of maximum radiation output. A Monte Carlo radiation transport simulation code, MCNP5, was used to design a proposed facility to shield two D-T neutron generators and one D-D neutron generator. A number of different designs were simulated, but due to cost and space issues a small concrete cave proved to be the best solution for the shielding problem. With this design, all of the neutron generators could be used and all of the rooms surrounding the neutron facility could be considered unrestricted to public access. To prevent unauthorized access into the restricted area of the neutron facility, light curtains, warning lights, door interlocks, and rope barriers will be built into the facility.