Quantification of various factors influencing the precision of thermoluminescent detector calibrations for new and used chip sets.

Factors affecting the random and systematic error in calibrating three sets of 100 LiF:Mg,Ti thermoluminescent detector chips were investigated. The chips were held in a polymethyl methacrylate plate with 0.3 cm deep wells covered with a thin top plate, affixed to a polymethyl methacrylate phantom 150 cm from a 3.2 × 10(10) GBq 137Cs source, used to irradiate the chips to 4.52 mGy. Three sets of chips were used: one new, one heavily used, and one having relatively high degrees of visible physical damage. Variations in the exposure rate across the plate were measured with an ion chamber. Experimental drift was judged by performing successive calibrations on subsequent days, while always reading the chips in the same order. The chips were subject to manual examination to determine variations in mass and physical quality. This study indicates that more accurate calibrations can be obtained by accounting for the error caused by nonuniformity in the delivered dose, which was in this study as high as 4.4% from the center to the edges of the target. Making use of more than three calibrations only reduces the standard deviation as a percentage of the mean of a set by less than 1%. Desirable dosimeters in commercially rejected sets were identified by comparing each dosimeter's standard deviation of response across all calibrations to a commercially acceptable control set. Up to 50% variations in mass and visual quality, including opacity, fracture, and surface scratches to chips, showed little to no correlation with their response.

Application of an equilibrium-based model for diffusion barrier charcoal canisters in a small volume non-steady state radon chamber.

Radon in indoor air is often measured using activated charcoal in canisters. These are generally calibrated using large, humidity- and temperature-controlled radon chambers capable of maintaining a constant radon concentration over several days. Reliable and reproducible chambers are expensive and may be difficult to create and maintain. This study characterizes a small radon chamber in which Rn gas is allowed to build up over a period of several days for use in charcoal canister calibration and educational demonstrations, as well as various radon experiments using charcoal canisters. Predictive models have been developed that accurately describe radon gas kinetics in the charcoal canisters. Three models are available for kinetics in the small chamber with and without radon-adsorbing charcoal canisters. Presented here are both theoretical and semi-empirical applications of this equilibrium-based model of radon adsorption as applied to canisters in the small chamber. Several charcoal canister experiments in the small chamber with an equilibrium-based model of radon adsorption applied are reported. Results show that it is necessary to include a continuous radon monitor in the chamber during canister exposures, as the radon removal rate is highly variable. Furthermore, the presence of the canisters significantly decreases the amount of radon in the small chamber, especially when several canisters are present. It was found that canister response in the small chamber is largely consistent with the equilibrium-based model for both applications, with average errors of 1% for the theoretical application and -4% for the semi-empirical approach.

Method of estimating lifetime cancer risk due to chronic radionuclide intake.

A methodology for computing the cancer risk due to chronic radionuclide intake, assuming that cancer risk functions per unit dose as a function of age are known, is presented. In this work, an age-dependent intake function is assumed, the total amount of activity present in the body at any given age is computed, and the annual dose equivalent or effective dose estimated using age-dependent dose conversion factors. In a series of time intervals extending from the age of intake to age 80 y, the radiation-induced cancer mortality is estimated by multiplying the dose in any given year by the cancer risk per unit dose at a given age. By integrating the product of the dose and the risk at each time interval, the overall risk due to various chronic radionuclide intake scenarios over a lifetime is determined. This result is compared to the risk computed using integrated committed dose quantities and to the risk computed from an age-independent risk per unit dose. The example cases of dietary contamination following a nuclear incident and uranium contamination in drinking water are presented. The results show that ignoring the age dependence of the dose-risk relationship underestimates the total lifetime risk by more than 80% for the dose due to ¹³7Cs in milk in a nuclear incident scenario. Furthermore, it is found that if the integrated committed dose quantity is used to evaluate risk, the total risk will be overestimated by almost 50% in the case of chronic uranium ingestion. These results demonstrate the sensitivity of the total lifetime risk to the proper assignment of dose to each time interval and to the use of age-dependent risk coefficients.

Simulation of a method for determining one-dimensional 137 Cs distribution using multiple gamma spectroscopic measurements with an adjustable cylindrical collimator and center shield.

With multiple in situ gamma spectroscopic measurements obtained with an adjustable cylindrical collimator and a circular shield, the arbitrary one-dimensional distribution of radioactive material can be determined. The detector responses are theoretically calculated, field measurements obtained, and a system of equations relating detector response to measurement geometry and activity distribution solved to estimate the distribution. This paper demonstrates the method by simulating multiple scenarios and providing analysis of the system conditioning.

An equilibrium-based model for measuring environmental radon using charcoal canisters.

Radon in indoor air is often measured using canisters of activated charcoal that function by adsorbing radon gas. The use of a diffusion barrier charcoal canister (DBCC) minimizes the effects of environmental humidity and extends the useful exposure time by several days. Many DBCC protocols model charcoal canisters as simple integrating detectors, which introduces errors due to the fact that radon uptake changes over the exposure period. Errors are compensated for by calculating a calibration factor that is nonlinear with respect to exposure time. This study involves the development and testing of an equilibrium-based model and corresponding measurement protocol that treats the charcoal canisters as a system coming into equilibrium with the surrounding radon environment. This model applies to both constant and temporally varying radon concentration situations, which was essential, as efforts are currently underway using a temporally varying radon chamber. It was found that the DBCCs equilibrate following the relationship E = (1 - e) where E is a measure of how close the DBCC is to equilibrium, t is exposure time, and q is the equilibration constant. This equilibration constant was empirically determined to be 0.019 h. The proposed model was tested in a blind test as well as compared with the currently accepted U.S. Environmental Protection Agency (U.S. EPA) model. Comparisons between the two methods showed a slight decrease in measurement error when using the equilibrium-based method as compared to the U.S. EPA method.

A numerical method for the calibration of in situ gamma ray spectroscopy systems.

High purity germanium in situ gamma ray spectroscopy systems are typically calibrated using pre-calculated tables and empirical formulas to estimate the response of a detector to an exponentially distributed source in a soil matrix. Although this method is effective, it has estimated uncertainties of 10-15%, is limited to only a restricted set of measurement scenarios, and the approach only applies to an exponentially distributed source. In addition, the only soil parameters that can be varied are density and moisture content, while soil attenuation properties are fixed. This paper presents a more flexible method for performing such calibrations. For this new method, a three- or four-dimensional analytical expression is derived that is a combination of a theoretical equation and experimentally measured data. Numerical methods are used to integrate this expression, which approximates the response of a detector to a large variety of source distributions within any soil, concrete, or other matrix. The calculation method is flexible enough to allow for the variation of multiple parameters, including media attenuation properties and the measurement geometry. The method could easily be adapted to horizontally non-uniform sources as well. Detector responses are calculated analytically and Monte Carlo radiation transport simulations are used to verify the results. Results indicate that the method adds an uncertainty of only approximately 5% to the other uncertainties typically associated with the calibration of a detector system.

Evaluation of total effective dose due to certain environmentally placed naturally occurring radioactive materials using a procedural adaptation of RESRAD code.

Due to a recent upward trend in the price of uranium and subsequent increased interest in uranium mining, accurate modeling of baseline dose from environmental sources of radioactivity is of increasing interest. Residual radioactivity model and code (RESRAD) is a program used to model environmental movement and calculate the dose due to the inhalation, ingestion, and exposure to radioactive materials following a placement. This paper presents a novel use of RESRAD for the calculation of dose from non-enhanced, or ancient, naturally occurring radioactive material (NORM). In order to use RESRAD to calculate the total effective dose (TED) due to ancient NORM, a procedural adaptation was developed to negate the effects of time progressive distribution of radioactive materials. A dose due to United States' average concentrations of uranium, actinium, and thorium series radionuclides was then calculated. For adults exposed in a residential setting and assumed to eat significant amounts of food grown in NORM concentrated areas, the annual dose due to national average NORM concentrations was 0.935 mSv y(-1). A set of environmental dose factors were calculated for simple estimation of dose from uranium, thorium, and actinium series radionuclides for various age groups and exposure scenarios as a function of elemental uranium and thorium activity concentrations in groundwater and soil. The values of these factors for uranium were lowest for an adult exposed in an industrial setting: 0.00476 microSv kg Bq(-1) y(-1) for soil and 0.00596 microSv m(3) Bq(-1) y(-1) for water (assuming a 1:1 234U:238U activity ratio in water). The uranium factors were highest for infants exposed in a residential setting and assumed to ingest food grown onsite: 34.8 microSv kg Bq(-1) y(-1) in soil and 13.0 microSv m(3) Bq(-1) y(-1) in water.

Calibration drift in a laboratory high purity germanium detector spectrometry system.

For unknown radionuclide identification, it is important that a high purity germanium (HPGe) spectrometry system be calibrated correctly for energy. The energy calibration of an HPGe spectrometry system will drift over time due to a variety of factors including the ambient temperature, the line voltage applied to the system, variation in the electronics, and other possible influences. In order to better understand the nature of this energy calibration drift, calibration spectra were collected over a period of several months from a laboratory HPGe spectrometry system. System parameters, including detector voltage, amplifier gain, and preamplifier gain, were not deliberately modified during the course of the experiment. The system was calibrated routinely over the 90 days, and the results of the calibrations were compared in order to assess the drift in the energy calibration of the detector over time. The analysis of a 36% high purity germanium system demonstrated the energy calibration drifted an average of 0.014 keV d(-1) to 0.041 keV d(-1) depending upon energy. At 1,332 keV, one day after calibration, it was shown that up to half of the total error in energy calibration was as a result of calibration drift.

The sunlight OSL response of a commercially available alpha-Al2O3:C personnel dosimetry material.

Carbon-doped, anion-defective aluminium oxide has become a widely used and effective medium for personnel dosimetry applications using optically stimulated luminescence (OSL) techniques. Though the commercial products currently using this material employ light-tight packaging to prevent light-induced effects on the OSL signal, the material could be employed in environments where package integrity cannot be assured. This paper reports on the results of an experiment performed to quantify the effects of sunlight exposure on alpha-Al2O3:C. Samples of commercially available Luxel material were exposed to carefully recorded levels of sunlight both before and after irradiations to determine the nature and magnitude of both activation and fading phenomena in this material. The results confirm that both fading and activation processes are seen in this material and indicate that the material reaches an equilibrium dose level in response to prolonged sunlight exposure equivalent to a dose of approximately 15 mGy under the experimental test conditions.

A simple radon chamber for educational use.

Radon chambers are typically able to maintain a constant, known concentration of radon by pumping a constant activity concentration of radon gas into the chamber. A radon chamber has been designed by placing a radon source inside the chamber volume and allowing radon to build up inside the chamber. Because the radon concentration is not constant, calculations have been made to determine the integrated equivalent constant radon concentration in MBqhm for up to 4 d. The chamber, made by placing a radium dial inside an incubator, has interior dimensions of 87.6 cm x 55.9 cm x 51.4 cm for a total volume of 0.25 m. The chamber can produce an integrated equivalent constant radon activity concentration level of 0.013 MBq h m over the initial 24 h, 0.043 MBq h m over the initial 48 h, 0.078 MBq h m over the initial 72 h, and 0.118 MBq h m over the initial 96 h. The chamber can also demonstrate, for educational purposes, the kinetics of the build-up of a radioactive gas in an enclosed environment as well as the kinetics of washout and leakage.

An evaluation of the Kearny Fallout Meter (KFM), a radiation detector constructed from commonly available household materials.

A radiation detector constructed of common household materials was developed at Oak Ridge National Laboratory (ORNL) by Cresson H. Kearny and has been referred to as the Kearny Fallout Meter (KFM). Developed during the height of the Cold War, the KFM was intended to place a radiation meter capable of measuring fallout from nuclear weapons in the hands of every U.S. citizen. Instructions for the construction of the meter, as well as information about radiation health effects, were developed in the form of multi-page newspaper insert. Subsequently, the sensitivity of the meter was refined by a high school teacher, Dr. Paul S. Lombardi, for use in demonstrations about radiation. The meter is currently being marketed for survivalists in light of potential radiation terrorist concerns. The KFM and Lombardi's variation of it are constructed and evaluated for this work. Calibrated tests of the response and variations in response are reported. A critique of the multi-page manual is made. In addition, the suitability of using such a detector, in terms of actual ease of construction and practical sensitivity, is discussed for its use in demonstrations and introductory classes on nuclear topics.

A comparison of transient dose model predictions and experimental measurements.

The RADTRAN and RISKIND transportation risk analysis computer codes are the primary tools used to estimate dose consequences and risks associated with the transport of radioactive material. Over the years, some of the mathematical models used within the two computer codes have been updated and the methodologies to calculate input parameters have been improved. In addition, both codes have been evaluated for ease of use and appropriateness of application and verified against other computer codes that perform similar calculations. However, neither code has been validated against experimental data. This report discusses the results of five sets of experimental measurements used to partially validate the specific mathematical models used to predict the dose to an individual due to a passing shipment of radioactive material within the RADTRAN and RISKIND computer codes. Based on the comparisons it was found that RISKIND most closely predicted the measured dose in the majority of the investigated scenarios and that 12 out of 14 cases demonstrate the expected inverse relationship between the measured dose and the distance of closest approach. Only half of the data demonstrated the expected inverse relationship between dose and speed of travel.

Numerical simulation of a TLD pulsed laser-heating scheme for determination of shallow dose and deep dose in low-LET radiation fields.

A new method is described to determine the depth-dose distribution in low-LET radiation fields using a thick thermoluminescent dosimeter (TLD) with a pulsed laser-heating scheme to obtain TL glow output. The computational simulation entails heat conduction and glow curve production processes. An iterative algorithm is used to obtain the dose distribution in the detector. The simulation results indicate that the method can predict the shallow and deep dose in various radiation fields with relative errors less than 20%.

Artifacts, anatomical and physiological variants, and unrelated diseases that might cause false-positive whole-body 131-I scans in patients with thyroid cancer.

The whole body 131-I scan remains an important component in the postoperative treatment of patients with well-differentiated thyroid cancer. Because normal thyroid tissue remnants and residual or metastatic foci of well-differentiated thyroid cancer have the unique ability to concentrate, organify, and store 131-I, the whole body scan provides a depiction of those tissues that can be ablated with therapeutic doses of 131-I. Over time, it has become obvious that the whole body scan may also reveal foci of 131-i uptake owing to a wide variety of other causes. We provide a detailed pathophysiological classification of the artifacts, anatomic and physiological variants, and nonthyroidal diseases that may give rise to false-positive whole body scans in postoperative patients with thyroid cancer. These include ectopic foci of normal thyroid tissue; nonthyroidal physiological sites (eg, choroid plexus, salivary glands, gastric mucosa, urinary tract); contamination by physiological sections; ectopic gastric mucosa; other gastrointestinal abnormalities; urinary tract abnormalities; mammary abnormalities; serous cavities and cysts; inflammation and infection; nonthyroidal neoplasms; and currently unexplained causes. This article also provides a detailed review of the widely scattered English language literature in which these phenomena were originally described.

Accounting for 222Rn loss during oven drying for the immediate laboratory gamma-ray spectroscopy of collected soil samples.

Drying soil samples in an oven to remove water alters the 222Rn emanation rate. Measurements of the oven drying 222Rn emanation rate from soil were made with a continuous radon monitor and the degree of 222Rn disequilibrium was quantified by laboratory gamma-ray spectroscopy. This paper presents a disequilibrium correction where the 226Ra activity in oven-dried soil samples is inferred from immediate laboratory gamma-ray spectroscopy of 214Bi before 222Rn and its decay progeny reach secular equilibrium.

Computer-based radiation safety training for hospital radiation workers.

Conducting a hospital-based radiation safety training class may lead to temporary technologist staffing shortages resulting in a reduction of patient services or even the cessation of all routine patient services. Use of an interactive computer-based radiation safety training software program may provide a practical alternative for hospital diagnostic and therapeutic radiation departments, as well as other hospital departments utilizing radiation sources, in meeting annual radiation safety training requirements for radiation workers. Medical radiation workers' participation in computer-based radiation safety training can make a positive impact on radiation safety awareness in the hospital, assist license holders in satisfying regulatory training requirements, ensure maximum participation of staff technologists, and reduce the burden of technologist staffing shortages caused by traditional methods of training.

A comparison of minimum detectable and proposed maximum allowable soil concentration cleanup levels for selected radionuclides.

Regulations on the release of a radioactively contaminated site for unrestricted use are currently being established by the Environmental Protection Agency. The effective dose equivalent rate limit for the reasonably maximally exposed individual was proposed at 0.15 mSv y(-1). The purpose of this study is to investigate whether or not maximum allowable soil concentrations of common radionuclides corresponding to 0.15 mSv y(-1) are readily detectable. These maximum allowable soil concentrations were estimated using RESRAD. The RESRAD estimates account for an effective dose equivalent rate from external radiation plus the committed effective dose equivalent rate from internal radiation delivering 0.15 mSv y(-1) to the reasonably maximally exposed individual. For Michigan and Arizona soil, the minimum detectable activities were calculated for a few radionuclides and compared to the RESRAD estimated maximum allowable concentrations. Considering only gamma-ray spectroscopy, this study found no evidence that concentrations of gamma-ray emitting radionuclides in soil contributing to 0.15 mSv y(-1) were undetectable.

Effects of intake function shape on internal dose estimations.

The estimated values of time-integrated dose equivalent are different if one assumes continuous intake, such as is the case for longer-term environmental exposures, instead of instantaneous intake, which is the assumption made for most standardized computational procedures. This paper presents the solutions of the first order kinetic equations for the ICRP 30 models of the respiratory system (RS) and the gastrointestinal tract (GI) for continuous activity intake. These solutions are applied to compute the cumulated number of disintegrations for different intake and integration times in order to show the effects of the assumptions of instantaneous intake vs. continuous intake on dose. For pulmonary clearance class D compounds instantaneous and continuous intake functions result in dose estimations agreeing to within 10% (when normalized to total intake) for integration times larger than 8 d for both RS and GI, for the case of radionuclides with radiological half-lives ranging between 8 d and 10(7) y. For 141Ce (class W) for both RS and GI the temporal intake function affects the dose by less than 10% for integration times greater than 280 d. In this case, the results for GI are less sensitive to the type of intake assumed than those for RS: the ratio of the number of transformations with the two intake assumptions has values of 0.8 or less for intake times of less than 7 d for GI and for intake times of less than 153 d for RS. For radionuclides such as 89Sr and 90Sr (class Y), the magnitude of the dose is very sensitive to the assumption made regarding the intake and has a strong dependence on the radionuclide half-life: for a 10% accuracy in dose evaluation for a continuous intake treated as instantaneous, the integration time must be greater than 650 d for 89Sr and 18,000 d for 90Sr. For all cases studied, the assumption of instantaneous intake overestimates the integrated dose for a continuous intake. More accurate internal dose evaluations may be made by consideration of the intake function shape when dose integration times less than approximately 10 y are of interest and non-instantaneous intakes are believed to have occurred.

Lost life expectancy rate: an application to environmental levels of radiation.

The risk index Lost Life Expectancy Rate (LLER) provides a unitless number (time of life expectancy lost per time exposed) describing the risk from exposure to a given hazard or from partaking in a given activity. Simple equations to calculate the LLER from radiation-induced cancers caused by an exposure to low-level radiation were derived using the relative risk models developed by the Nuclear Regulatory Commission (upper bound estimate), the United Nations Scientific Committee on the Effects of Atomic Radiation-1988, and the National Academy of Science's Committee on the Biological Effects of Ionizing Radiation (BEIR V). Estimates of the LLER to an average person from a continuous exposure to 0.1 microSv h(-1) based on these models, respectively, are 5.5 x 10(-4), 9.5 x 10(-4), and 9.4 x 10(-4). These values compare to LLERs of 0.015 from occupational accidents, 0.25 from being an automobile passenger, and 2.0 from cigarette smoking. Factors effecting LLER from radiation exposures examined in this work include dose rate, age, sex, race and smoking status.

Comparison of in situ and laboratory gamma spectroscopy of natural radionuclides in desert soil.

In situ and laboratory gamma spectroscopy were used to characterize natural background levels of radiation in the soil at eight sites around the Yucca Mountain Range. The purpose of this practical field analysis was to determine if published empirical in situ calibration factors would yield accurate quantitative specific activities (Bq kg(-1)) in a desert environment. Corrections were made to the in situ calibration factors to account for the on-axis response of a detector with a thin beryllium end window. The in situ gamma spectroscopy results were compared to laboratory gamma spectroscopy of soil samples gathered from each site. Five natural radionuclides were considered: 40K, 214Pb, 214Bi, 208Tl, and 228Ac. The in situ determined specific activities were consistently within +/-15% of the laboratory soil sample results. A quantitative discussion of the factors contributing to the uncertainty in the in situ and laboratory results is included. Analysis on the specific activity data using statistical hypothesis tests determined that three nuclides, 214Pb, 214Bi, and 228Ac showed a weak site dependence while the other two nuclides, 40K and 208Tl, did not exhibit a site dependence. Differing radiation background levels from site to site along with in situ and laboratory uncertainties in excess of 10% are two factors that account for the weak site dependence. Despite the good correlation between data, it was recommended that the in situ detector be calibrated by a detector-specific Monte Carlo code which would accurately model more complex geometries and source distributions.