The 54th Annual Meeting of the Health Physics Society
July 12-16, 2009
Minneapolis, MN

Single Session

[Schedule Grid]

P - Poster Session

Room: Exhibit Hall A   1:00 - 3:00 PM



P.1   Detection Limit as a Function of Electron Energy for Delayed Neutron Yields from Photofission of U-238 M Ankrah*, Idaho State University, Pocatello ; KC Chandler, Idaho State University, Pocatello; A Hunt, Idaho State University, Pocatello

Abstract: Some active interrogation methods which use photon (bremsstrahlung) sources from accelerators have been used in national security and nuclear nonproliferation applications. Research and technology developments for special nuclear material (SNM) detection have grown due the possibility of terrorist attack using a nuclear weapon. The detection of radiation signals from concealed SNM poses a challenge since shielding alters the energy spectrum of these signals. This study used bremsstrahlung photons with end point energies of 6, 10, 14, 18 and 22 MeV to induce fission in U-238 with subsequent measurement of the delayed neutron yields using He-3 photoneutron detectors. For each of the energies selected, the measured dose delivered on target was used to normalize the delayed neutron yield. The delayed neutron yields in the presence of polyethylene, lead and steel is presented.


P.2   The Evaluation of Symptom Ringing Disillusion Among Children and Adult Cellular Telephone Users N. Kumar*, Babasaheb Bhimrao Ambedkar University, Lucknow, India ; V. P. Sharma, Indian Institute of Toxicology Research, Lucknow, India; N. Mathur, Indian Institute of Toxicology Research, Lucknow, India; M. Y. Khan, Babasaheb Bhimrao Ambedkar University, Lucknow, India; R. A. Khan, Babasaheb Bhimrao Ambedkar University, Lucknow, India

Abstract: Cellular technology has been a blessing to all human beings for the last decades. More than 6 billion cellular telephone users are holding the device in the whole world and day to day utilization period of cellular telephone are increasing among the users. Presently children are more fascinated to make use of the cellular telephones and in future they would be exposed extensively to radiofrequency radiation (RFR) for a long period and after passing 20-30 years this generation would be consider as highest exposed group of cellular irradiation. The developing phase of the childrens brain and its hypersensitivity to RFR tends to concern about the cellular telephone safety particular relevance to central nervous system functioning. Ringing disillusion may consider as psychological abnormality in which a cellular telephone user feels suddenly ringing voice of the device but in actually there is no ringing voice ranged at that time. In the study an assessment have been made about ringing disillusion i.e. common psychological abnormality among children and adult cellular telephone users. In the study we assessed ringing disillusion among normal users (NU<500 hours life time exposure) and heavy users (HU>500 hrs life time exposure) in both children (users > 20 years age) and adult (user > 20 years age) telephone users. A survey study was conducted among the randomly selected 188 cellular telephone users in Lucknow city of India. In the study subjects were asked for their demographic and socio-economic characteristics utilization pattern of the cellular telephone adopted safety measures during call per day utilization and year of utilization of the device. Surveyed data of children (< 20 years age) and adults (> 20 years age) were separated and classified to normal user (NU<500 hours life time exposure) and higher user (HU>500 hrs life time exposure) according to life time exposure of cellular telephone (i.e. per day utilization X number of utilization years). Data were analyzed and comparison was made individually between NU and HU in both groups. In this study normal users were 26.32% and heavy users were 25.95% they reported ringing disillusion. Children and adult cellular telephone users were reported ringing disillusion 11.11% and 27.65% respectively. Among the children 14.28% subjects were found normal user and 9.09% subjects were heavy user. However in adult users,normal users and heavy users were associated 28% and 27.5% to ringing disillusion. Analyzed data suggested that adult normal user have twice sensitivity to ringing disillusion in comparison to children normal user. But this sensitivity was thrice among adult heavy users in respect to children heavy user. However this ratio could not attain statistical significant because of small number of subjects in adult user. The association of the ringing disillusion among the adults is higher in comparison to children but symptom does not link to life time use of cellular telephone. There is need the clinical studies over the mechanisms whereby cellular telephone irradiation may alter psychological functions of biological systems. Millions of people are at blurred that cellular phone irradiation is truly unsafe. The actions are required over the concept of protecting our health and environment from RFR exposures that is probably but not definitely unsafe.

P.3   Effects of Ionizing Radiation Exposure on Arabidopsis thaliana T Kurimoto*, German Cancer Research Center (DKFZ) ; J. V. H. Constable, California State University, Fresno; A Huda, California State University, Fresno

Abstract: The effects of ionizing radiation on Arabidopsis thaliana were investigated using 6MV x-ray produced by a linear accelerator. Photosynthesis and respiration rates, chlorophyll fluorescence fv/fm ratio and yield, plant height, total leaf area, stem mass, leaf mass, and above-ground biomass were measured in order to evaluate both physiological and physical impacts of ionizing radiation. The statistical analysis of the radiation effects with respect to four different total doses (0.5Gy, 5Gy, 50Gy, and 150Gy), two different treatment types (single and fractionated), and three different life stages at irradiation (15-day, 20-day, and 25-day old) revealed the details of dynamic responses of Arabidopsis thaliana to radiation exposure. The results illustrated that the age at the time of irradiation was a key to determine the overall effects of radiation exposure, and the irradiated Arabidopsis indicated much greater contrast in physical growth compared to physiological responses immediately after the irradiation. The investigation suggests that plants like Arabidopsis are capable of being utilized as a biodosimeter and further studies can be performed on specific areas in order to assess the effects of ionizing radiation for a practical application.


P.4   Evaluation of Innovative Technology for Decontamination of Contaminated Surfaces SK Dua*, FIU ; L Lagos, FIU; D Calderin, FIU; M Ngachin, FIU; RA Colon Mendoza, FIU

Abstract: An innovative technology developed by the Y-12 National Security Complex has been identified for evaluation for decontamination of a number of surface media used in nuclear industry. It comprised of a cloth-like special wiping material, which upon wiping attracts dust particles dislodged from the surface by van der Waals forces, The van der Waals forces between the wiping material and contaminant particles inhibit release of airborne particles into the atmosphere. The surfaces selected for this evaluation are: Formica, stainless steel, coated carbon steel, coated concrete, tiles and Plexiglas. The technology evaluation process comprises of: 1) Contaminating the surfaces with known radioactivity concentration of short-lived beta emitters such as P-32, P-33 and S-35, and measuring contamination on surfaces before and after wiping; 2) Measuring residual contamination, mainly Radium 226, on stainless steel glove box internals before and after wiping the surfaces. The glove box housed a loop, which was previously used for rheological and hydrological studies of Fernald silo materials. The loop and other equipment were dismantled and the glove box decontaminated using the baseline technology; however, small amounts of contamination remained on glove box internals and therefore served as a test pad for evaluation of this decontamination technology. 3) Spraying a fluorescent powder on the surfaces and making qualitative assessments of the amount of the material, using UV light, before and after wiping the surface. As a part of this technology evaluation, to determine the quantity of material that becomes airborne in all above described demonstrations, air samples are collected during the entire period of wiping and are continued 15 minutes after completion of wiping. Air samples are assessed for the material collected on air sample filters. Comparison of innovative technology is made with baseline technology of wiping with rags both for removal of contamination from the surfaces and for the material that becomes airborne.

P.5   Popularization of Science in the Nuclear Area Focusing on Stakeholders Living Nearby Decommissioned Uranium Mines Fabiana Dias*, Brazilian Nuclear Energy Commission ; Taddei M.H.T., Brazilian Nuclear Energy Commission; Mello del Britto Edenil, Brazilian Nuclear Industries; de Azevdo Py Junior Delcy, Brazillian Nuclear Industries

Abstract: Nuclear energy is a theme that involves a lot of controversy, ignorance, fear, and also hopes. The expansion of the use of nuclear energy faces resistance from many segments of society. One characteristic of the Brazilian Nuclear Program is that ores as well as nuclear facilities are located near populations in need of technological information. This frequently generates unfounded resistance in the population towards nuclear energy projects, aggravated by opinions of low credibility of regulatory agencies and broad coverage by the media. The objective of this paper is to present a project focused on the Poços de Caldas Plateau in Brazil, which is known for its elevated natural radioactivity due to the presence of radioactive ores (a region with approximately 280,000 inhabitants). The region has a uranium mine that is undergoing decommissioning and closure. The project uses three scientific approaches to evaluate the problem, utilizes a variety of public information tools and considers use of internationally accepted demographic parameters. This approach has made it possible for a qualified multi-disciplinary team to make technically, economically and politically viable actions to continue with the site decommissioning and closure projects. The project’s scientific, technically safe, and ethically reliable information is made available in a planned and systematic manner in order to improve scientific education of Stakeholders. Such information includes: effects of radiation on the health of the local population and the environment, levels of radiation dosage, magnitude of doses, cancer epidemiology, technological uses of nuclear energy, benefits and damages of radiation, as well as other information. Implementation of this project, in both short and long term, contributes to an improvement in the scientific education of the population, resulting in a change of paradigm and creating a capacity for interaction and dialog between communicators and identified Stakeholders.

P.6   Radioecological Criteria and Norms During Remediation of Shore Infrastructure of Nuclear Fleet Nataliya Shandala, Burnasyan Federal Medical Biophysical Centre, Moscow ; Vladimir Seregin*, Burnasyan Federal Medical Biophysical Centre, Moscow; Malgorzata Sneve, Norwegian Radiation Protection Authority, Oslo; Alex Titov, Burnasyan Federal Medical Biophysical Centre, Moscow; Dmitriy Isaev, Burnasyan Federal Medical Biophysical Centre, Moscow

Abstract: Remediation of sites and facilities belonging to the sites of temporary storage (STS) of spent nuclear fuel (SNF) and radioactive wastes (RW) at Andreeva Bay and Gremikha village on the Kola Peninsula is one of the regulatory functions of radiation protection. After termination of operation of the ex-Navy serviced facilities, their infrastructure degraded resulting in radioactive contamination of some parts of the site. As one end-goal of remediation is putting STS into ecologically safe conditions, elaboration of quantitative radiation-ecology criteria and norms for STS site and facility conditions is urgent. Remediation criteria and norms defining requirements for radiation protection of workers, public and limits of environmental contamination have been developed for three main options: conservation, conversion (partial or overall renovation) and liquidation in the form of: dose limits and dose constraints; levels of radioactive superficial contamination of workshops and equipment inside them; specific activity of radionuclides in marine media, including fish; activity concentration of radionuclides in ground waters on-site STS. Dose reduction up to and lower adopted levels must use optimization principle accounting the particular use of remedied sites and facilities. Established dose constraints and contamination levels are the highest at optimization. The resulted optimized dose values serve as a basis for determination of specific levels of the site and facility cleanup. Results of radiation monitoring and control must confirm observance of developed norms and criteria during and after remedial actions.


P.7   Review of Radon Assessment Studies In The City Of Tbilisi, Republic Of Georgia Samson Pagava, Tbilisi State University ; Vladimer Rusetski, Tbilisi State University; Gouram Kutelia, Tbilisi State University; Nikoloz Shubitidze, Tbilisi State University; Roy Dunker, Idaho State University; Eduardo Farfan, Savannah River National Laboratory; James Popp, York College of the City University of New York; Jason Harris*, Idaho State University; Doug Wells, Idaho State University; Maia Avtandilashvili, Idaho State University

Abstract: Since 2007, Georgian and U.S. researchers have been carrying out radon (Rn-222) assessment studies in the city of Tbilisi, located in the Republic of Georgia. Churches were initially selected for testing due to their wide-spread distribution and relatively secure working environment. When permission was granted, dwellings and offices were also tested. At the test sites, radon testing in soil, air and water were carried out, supplemented with background gamma measurements. Measurements were taken with E-PERMs, RAD7, and GammaTRACER instruments. At the densely populated Tbilisi administrative districts of Old Tbilisi and Vake-Saburtalo, the following results were obtained: radon concentrations in soil ranged from 0.1 to 19 kBq per cubic m; gamma dose rates ranged from 80 to 160 nSv/h; radon concentrations in water ranged from 2.0 to 8.5 Bq/L; and radon concentrations in air measured up to ~700 Bq per cubic m. The indoor radon concentrations at several test sites exceeded the regulation levels adopted by the United States, European Union and the Republic of Georgia. The project results were compiled in a web-accessible database, ( for governmental, scientific and public use. Georgian and English versions of an information pamphlet explaining radon and this project were created, and distributed to residents of Tbilisi. Project results were communicated to the Georgian Ministry of Labor, Health and Social Affairs. The armed conflict between Russia and Georgia in August 2008 caused significant delays and logistical problems for the Georgian project team members. Complete results and conclusions of the study are presented.

P.8   Radiation Situation Nearby the Uranium Mining Facility Nataliya Shandala, Burnasyan Federal Medical Biological Centre, Moscow ; Anna Filonova*, Burnasyan Federal Medical Biological Centre, Moscow; Alex Titov, Burnasyan Federal Medical Biological Centre, Moscow; Dmitriy Isaev, Burnasyan Federal Medical Biological Centre, Moscow; Vladimir Seregin, Burnasyan Federal Medical Biological Centre, Moscow; Maria Semenova, Burnasyan Federal Medical Biological Centre, Moscow; E.G. Metlyaev, Federal Medical Biophysical Centre, Moscow, Russia

Abstract: Octyabrsky village is located in uranium- and radon-containing biogeochemical province, near tectonic break in Transbaikalia region of Western Siberia, on the Russian-Chinese boundary. It is situated within the industrial site of the uranium mining facility, i.e., in the health protection zone of the mining chemical production association. Considerable radon inflow in dwellings was firstly detected in 1987. Radon concentrations registered in flats were comparable with its air concentration in underground openings of mining. In some rooms, radon air concentration acceptable for the personnel category À (miners involved into underground operations) was excess. Gamma dose rates currently observing within the village are specific for uranium-containing areas. Inside dwellings, within farmlands and outdoors they are 0.09–0.4, 0.2–0.8 and 0.1–2.5 mcSv/h, respectively. The man-made contamination change of the village top-soil is different, depending upon natural radionuclide composition. On the most part of Octyabrsky area, natural radionuclide composition specific activity, particularly 226Ra content in soil, is higher a bit than in the controlled area. In some streets, 226Ra specific activity in soil is 200 – 400 Bq/kg. 226Ra and 210Pb specific activity in local foods (milk, potato, vegetables)sampled in Octyabrsky is higher (in milk - up to 10 times), in comparison with the controlled village. Radon EEAC in 40% dwellings exceeds 200 Bq/m3acceptable level. According to the sanitary norms, protective actions are to be implemented here to reduce radon content - cellar backfilling, floor weatherproofing, and overhaul of houses. However, these measures appeared to be inefficient: they did not cause reduction of radon inflow in dwellings. So, a provision was issued prohibiting public living within health protection zone of mining facility. Excess norms on radon are a case in favor of the village resettlement. Arrangement and performance of remedial measures of uranium mining sites will need in future.

P.9   Effective Method for Simulation of the Radioactive Material Dispersion in Terrestrial Surface Water Bodies Wen-hwa Wang, Institute of Nuclear Energy Research, Atomic Energy Council ; Jeng-Jong Wang*, Institute of Nuclear Energy Research, Atomic Energy Council; Bor-jing Chang, Institute of Nuclear Energy Research, Atomic Energy Council; Ing-jane Chen, Institute of Nuclear Energy Research, Atomic Energy Council

Abstract: In the past most of the studies on radioactive dispersion device (RDD) are focused on the model of air dispersion. The purpose of this paper is to provide an effective approaches to simulate the complex 3D dispersion and transportation process of radioactive contaminants in irregular flow environment. The simulation methods included in this report are based on the combination of cumulative discharge method and finite element method. The critical points in water resources protection are the estimation of the concentrations for the surface water and sediments. The major procedures to calculate the concentrations are as follows: describing the geometry of the domain, entering the coordinates of each node, selecting parameters for each node of the mesh, defining velocity and lever of the water body, entering dispersion characteristics & given values, solving finite element method equations, deriving the stream function and output to files. Finally, the solution of advection-dispersion equation is superposed on the streamlines. In conclusion, the approaches of this study combine the application of partial differential equations, fluid dynamics, hydrology and a user-friendly computer code. This 3D simulation method is valid for both of short release and continuous release at any location of surface water. The validation result shows that the contaminant concentrations derived from this study and that from the EPA approved COMIX software are perfect mach.

P.10   A STUDY OF STRONIUM 90 ANALYSIS METHOD BY LIQUID SCINTILLATION COUNTING FOR THE ENVIRONMENTAL SAMPLES Jeng-Jong Wang*, Institute of Nuclear Energy Research, Atomic Energy Council ; Hsin-Fa Fang, Institute of Nuclear Energy Research, Atomic Energy Council

Abstract: The radioactivity of strontium 90 containing in the environmental sample is determined by chemically purified with a crown-ether resin column, and then, after mixing with the organic scintillation cocktail, measured directly by a liquid scintillation counter without waiting for the growth of yttrium 90. This method is verified with seven kinds of environmental samples, including water, filter, soil, tea leaf, milk powder and meat, which are spiked with standard source of strontium 90 for reference, and the performance is estimated. The estimated results reveal that, the counting efficiencies of strontium 90 and yttrium 90 are larger than 90%, and the detection limit is lower than 0.04 Bq. Furthermore, the bias between the measurement results and the reference values are less then 2.0%, and the standard variation of the measurement values are less then 2.1%.

P.11   DEVELOPMENT OF THE ENVIRONMENTAL GAMMA MONITORING NETWORK FOR EMERGENCY RESPONSE PURPOSES IN TAIWAN M. C. Horng*, Radiation Monitoring Center, AEC ; F. C. Huang, Radiation Monitoring Center, AEC; M. F. Kao, Radiation Monitoring Center, AEC; C. C. Liu, Radiation Monitoring Center, AEC; H. H. Tseng, Institute of Nuclear Research, AEC ; C. C. Huang, Radiation Monitoring Center, AEC

Abstract: An environmental gamma-monitoring network in Taiwan, established in 1990, has an important place in determining the airborne gamma doses and in altering the governmental authorities when an unexpected release has been detected. In recent years, the network has upgraded the monitoring and communication hardware to create a smart configuration that fits the requirements for emergency response purposes. An effective approach has been developed to make rapid and accurate decisions based on real time, available and historical data from all monitoring stations and units. As of December 2008, this network is equipped with real time instrument installed at twenty-eight sampling stations, located at around the nuclear facilities, populated cities, as well as offshore islands of the Taiwan area. With round the clock real time monitoring, data are transmitted through the network and real time monitoring information is provides via RMC's website to the public with transparency. The monitoring results during the whole year show that all station is within the variation of environmental background radiation. Moreover, several of the instrument units in the network have been built for mobile and portable applications, and are equipped with GPS (global positioning system) and 3G (third generation radio card) to enhance emergency response capabilities. These units are available for the measurement of gamma dose rate and nuclide-specific during the 2008 nuclear emergency exercise in southern Taiwan.

External Dosimetry

P.12   Investigation of a Model for the Fading of Thermoluminescent Dosimeter Glow Curve Peak Areas in the Presence of Chronic Irradiation J. A. Harvey*, University of Michigan, Ann Arbor ; E. M. Thomas, University of Michigan, Ann Arbor; K. J. Kearfott, University of Michigan, Ann Arbor

Abstract: Thermoluminescent dosimeters (TLDs) are continuously exposed to radiation and their signals undergo fading during the time between annealing and readout. The resulting sensitivity variations as a function of time for the respective glow curve peaks have been previously characterized as a combination of pre-irradiation sensitivity changes and post-irradiation fading, with irradiation viewed strictly as a single acute event with background irradiation being neglected. A more in-depth analysis addresses the fading during readout, the addition of background, and the affects of both pre-irradiation fading and post-irradiation fading. A reasonable approach would be to allow TLDs to sit for known times at background radiation, subject them to significant acute irradiation, then immediately read them out and perform a peak analysis on the resulting glow curve. This would allow characterization of a pre-irradiation sensitivity function valid at background dose rates. TLDs could then receive a significant acutely delivered dose immediately post-annealing, and allowed to sit at background with various amounts of time post-irradiation to derive a post-irradiation fading function appropriate to background dose rate levels. The temporal integration of the product of these two functions and dose during the deployment time would then yield a model of the total signal for a given glow curve peak when the dose of interest is delivered acutely. It is not clear that different processes are involved in TLD sensitivity changes and signal fading before and after irradiation, nor for acute and chronically delivered doses. An improved model would be to derive an empirical expression that would be applicable for the sensitivity changes that occur during dosimeter deployment times in general. Varying doses could be delivered as different functions of time in order to validate such a model.

P.13   Comparison of Peak-Determined Region of Interest and Glow Curve Peak Fitting Analysis of Thermoluminescent Dosimeter Data E. M. Thomas*, University of Michigan, Ann Arbor ; J. A. Harvey, University of Michigan, Ann Arbor; B. M. Wu, University of Michigan, Ann Arbor; K. J. Kearfott, University of Michigan, Ann Arbor

Abstract: There are several different methods for determining the dose measured by a thermoluminescent dosimeter (TLD) for which the area under the glow curve is assumed to be proportional to dose. One approach is to set a region of interest (ROI) in the glow curve and to take the area under the curve within the ROI. Another method is to run a peak fitting program and consider the area as computed from the fitted parameters describing these peaks. To collect the data for a comparison of the two analysis approaches and to assess the relative precisions and accuracies across a set of TLDs, a group of 100 LiF:Mg,Ti TLD chips were irradiated by a 3.4 x1011 Bq 137Cs source to an average dose of 5 mGy. The TLDs were subsequently read on a commercial hot planchet TLD reader 30 min after irradiation. An ROI for each glow curve, running from the start of the TLD readout to the end of peak 5, was chosen and the area under the curve in this region calculated. Each glow curve was then analyzed with a peak-fitting program written in mathematical analysis software. Corrections for positional variation in dose and drift over the experiment, which consisted of ten irradiations, were applied. The area of the ROI and the area under peaks 2 to 5, as calculated by the fitting program, were compared. Individual trials for a TLD were normalized to the mean of all trials for the TLD, to evaluate the relative precision of the two methods. Preliminary results show that the mean standard deviation of these normalized trials was 5.07% for the peak fit method and 5.78% for the ROI method. This suggests that the peak fit method is slightly more precise than the ROI method.

P.14   Reproducibility of Glow Peak Fading Characteristics of Thermoluminescent Dosimeters B. M. Wu*, University of Michigan, Ann Arbor ; J. A. Harvey, University of Michigan, Ann Arbor; E. M. Thomas, University of Michigan, Ann Arbor; R. J. Bergen, University of Michigan, Ann Arbor; S. E. Carney, University of Michigan, Ann Arbor; J. P. Newton, University of Michigan, Ann Arbor; K. J. Kearfott, University of Michigan, Ann Arbor

Abstract: A thermoluminescent dosimeter (TLD) is a small chip of ceramic material whose electrons are promoted to energetic states upon exposure to ionizing radiation and may then be trapped in intermediate states. Radiation dose is determined by measuring the amount of light emitted when heating de-excites the trapped electrons. Over time, the electrons in their excited states can drop back to their ground state on their own, decreasing the emitted light in a process called fading. While some changes have been observed to occur in sensitivity as a function of time pre-irradiation, fading after irradiation was the focus of this work. A set of 100 calibrated LiF:Mg,Ti TLDs of dimensions 3.2 × 3.2 × 0.9 mm3 was first annealed. The set was then irradiated to an average of 5 mGy using a 3.3 × 1011 Bq 137Cs source. The irradiated TLDs were allowed to fade at room temperature for 0.5, 1, 1.5, 2, 3, 4, 8, or 15 days. After each fading period, the TLDs were read out over a 3.5 h time period using a standard commercial hot planchet TLD reader with nitrogen gas to prevent chemiluminescence. A mathematics software package was used to fit the glow curve data to a five-peak first-order Gaussian kinetics model. The areas of peaks 2, 3, and 4 were then normalized to peak 5, the slowest-fading peak. The peak area ratios were then fit to single decaying exponentials that allowed the determination of the fading rate by calculating their parameters. Preliminary analysis of the data shows a fading rate of 0.221 ± 13%/d for peak 2, 0.035 ± 8.3%/d for peak 3, and 0.0089 ± 48%/d for peak 4. Maximum chip to chip variations in fading rates were 384% for peak 2, 355% for peak 3 and 332% for peak 4. In the group of 100 TLDs, approximately 7 to 10 TLDs had fading rates for peaks 2 to 5 that varied more than 1.5 standard deviations from the mean for the group.


P.15   Wireless Encrypted Ionizing Radiation Monitoring In Cargo/Port areas J Baumbaugh*, SSC-Pacific ; R Clement, SSC-Pacific

Abstract: In 2007, the Department of Homeland Security (DHS) released technical requirements for container monitoring/communications technology to protect our nation’s ports against criminal activity/terrorist attacks. Hundreds of millions of cargo containers enter land/sea-based ports daily. Although some port personnel are equipped with monitoring devices, hundreds of thousands of workers have potential to be exposed to toxic substances/ radioactive material (RAM) shipments entering the US, legitimate and not. Per the DHS requirements, facilities world-wide will be tracking cargo utilizing wireless sensors linked with network access points situated within harbors forming a wireless “cloud” of coverage. Encrypted GPS/inventory/security data will be exchanged as cargo is transported in/out of port. Readers receive encrypted binary data from various types of container monitors. Data will be transmitted via the internet to government/commercial data centers throughout the world. Equipping port-based network access-points with spectroscopic portal sensors to detect/identify RAM allow port security to monitor cargo BEFORE it reaches land, decreasing the likelihood of personnel radiation exposure. Portal sensors will give advanced warning to harbor/port security that RAM is entering port and give time to intercept the cargo before it’s unloaded from a vessel and provide information on specific radioisotope(s) and exposure rates at various points within the harbor. This paper describes a novel concept for utilizing this new port wireless infrastructure to support fully automated real-time monitoring of radiation species and exposure levels at port facilities for personnel safety and purposes of national security. Operational concepts, system architecture and implementation are discussed as well as ideas on sensor types and data content for the proposed automated monitoring systems.

P.16   Alpha 7L Alarm Set Points and Response Times DL Wannigman*, Los Alamos National Lab ; AT Martinez, Los Alamos National Lab

Abstract: Using the latest software version and simulating airborne radioactivity releases using electroplated sources, Los Alamos National Laboratory has optimized the performance of their Alpha 7L continuous air monitoring (CAM) system. LANL uses Alpha 7L CAMs to monitor for Pu-239, Pu-238/Am-241 and Np-237 with fast alarm set points as low as 1200 DAC and 12 DAC-h and slow alarm set points of 6 DAC and 3 DAC-h. Using a Pu-239 electroplated source masked down to an equivalence of 10 DAC-h, these CAMs reliably alarm in 30-60 seconds. Preliminary field experience reveals that 200 Alpha 7L CAMs equipped with these settings can be operated in LANL’s Plutonium Facility for 90 days without a false alarm. This equates to a false alarm rate of less than 0.02 false alarms per CAM per year.

P.17   Monte Carlo Spectral Simulations as Microcalorimeter Gamma-Spectrometer Design Tool E S*, Colorado State University ; T Johnson, Colorado State University; M Rabin, Los Alamos National Lab; J Ullom, National Institute of Standards and Technology

Abstract: Simulations are an ideal design tool for researchers working with novel technologies. This presentation describes a Monte Carlo simulation of Transition Edge Sensor (TES) microcalorimeter gamma-spectrometer Plutonium (Pu) spectra. The simulations predict the relative error in intensity measurements of Pu photopeak gamma-ray lines as a function of the spectrometer’s spectral resolution and system efficiency. The complexity of the Pu spectrum in the 94 to 105 keV region of interest demands a spectrometer of high spectral resolution to accurately and precisely measure the Pu lines and derive Pu isotopic ratios. Minor differences in Pu isotopic ratios yield significant insight to the history and potential use of Pu under assay and drive the desire for high accuracy and precision in spectral measurements. The spectral resolution of a single-pixel composite TES gamma-spectrometer is proportional to the square root of the volume of its photon absorber, while its efficiency is directly proportional to the volume. Hence, there is an intrinsic tension between resolution and system efficiency. The relative values of resolution and efficiency merit investigation to achieve optimal spectrometer design, and in this project, Monte Carlo spectral simulations were used to compare the relative measurement error in gamma-spectrometer systems with different combinations of resolution and efficiency. Simulated Pu spectra included overlapping Voigt-shaped x-ray and Gaussian-shaped gamma-ray photopeak lines. Line intensities, centroids, and a constant background were extracted from measured spectra or known nuclear data so as to generate authentic simulations. Previous analytical methods cannot predict measurement error in the case of more than two overlapping lines, or in the case of non-Gaussian line forms. The Monte-Carlo simulations described here are a powerful tool for assessing the relative value of spectral resolution and efficiency and for determining the most effective design of a microcalorimeter spectrometer.

Emergency Planning and Response

P.18   Establishing the mobile environmental survey system for radiological emergency Hsin-Fa Fang*, Institute of Nuclear Energy Research, Taiwan

Abstract: After nuclear power plant accidents at Three Mile Island and Chernobyl, Taiwan has intensified the efforts on the works of radiological emergency response and preparedness as other countries. The 911 event in 2001 made us realize that the environmental survey for radiological emergency should be more mobile than conventional methods for responding the uncertain places and occasions of radiological emergency. Therefore Institute of Nuclear Energy Research launched the project for developing modern tools used in radiological emergency response for Taiwan, including the establishment of a mobile environmental survey system. The considerations of establishing the system are base on what we want to improve and the conformability of routine environmental radiation monitoring. For purpose of the conformability of routine environmental radiation monitoring, we need use the instruments used in the routine monitoring for emergency survey. The emergency survey conforms to routine environmental radiation monitoring has many benefits including cost, maintainability of instruments, comparability of survey data and so on.


Abstract: The US Environmental Protection Agency (EPA) is working to assess the impact of an urban radiological dispersion device (RDD) and to develop containment and decontamination strategies for the recovery process. It is necessary to understand the interactions between urban surfaces and RDD radionuclides to develop optimized strategies that minimize the decontamination efforts due to the spread of the radioactive contamination and its penetration into, as well as its binding to, urban surface materials. RDD decontamination activities will likely commence weeks to months after an event; therefore the contaminated area will experience a variety of weather conditions (rain, snow, relative humidity (RH) variation, etc). This may allow penetration and reaction of water soluble radioactive materials, such as cesium chloride, into permeable surfaces, thereby increasing the difficulty of their removal. This delay also allows time for these radioactive materials to bind with the urban surfaces. In this work, several analytical methods including laser ablation - inductively coupled plasma - mass spectrometry (LA-ICP-MS), traditional ICP-MS, and X-ray photoelectron spectroscopy (XPS) were used to study the binding of cesium chloride (133CsCl) and its migration into four urban materials; limestone, fine aggregate concrete, coarse aggregate concrete, and red brick, as a function of relative humidity. Physicochemical interactions also affect how efficiently cesium can be removed from porous urban surfaces using gross decontamination methods such as water wash-down. Therefore, related studies are presented which assess the issues surrounding the use of water from fire hoses to decontaminate porous urban surfaces contaminated with cesium chloride. In these laboratory scale studies, a typical fire hose incident water pressure and flow were simulated to evaluate the decontamination efficacy and fate of the cesium. Deposition mechanism (liquid spray, dry particulate deposition, and dry particulate deposition at high RH) and urban material type (fine aggregate concrete, limestone, and brick), were varied to determine their effects on water-based decontamination efficacy and the fate of cesium. Results are discussed along with estimates of wastewater volume generated during decontamination.

P.20   Interregional training of radiation emergency medical assistance for developing countries - Experience of Burnasyan Federal Medical Biophysical Center A. U. Bushmanov*, FMBC of FMBA of Russia ; K. V. Kotenko, FMBC of FMBA of Russia; A. S. Kretov, FMBC of FMBA of Russia; V. I. Krasnuk, FMBC of FMBA of Russia

Abstract: Burnasyan Federal Medical Biophysical Center of the Federal Medical Biological Agency of Russia is known for its experience and high-level expertise. From clinical records it has the following estimation of treated cases per year: • acute radiation sickness and patients with leukaemia exposed up to 12 Gy - 5-6 cases • local radiation injury – 1- 2 cases Experience shows that developing countries using different sources of ionizing irradiation, but have not enough qualified specialists in radiation medicine to organize medical assistance in case of radiation injures. So, Burnasyan Federal Medical Biophysical Center providing fellowship for participants from developing countries: relevant medical specialists (haematologists, specialists on occupational diseases, radiopathologists, surgeons in burn departments etc.), potentially involved in treatment of overexposed, injured or contaminated patients. The aim of the fellowship is to provide medical personnel involved in a medical response to radiation emergencies with practical knowledge and experience in treatment and management of acute radiation sickness and local radiation injuries. The main activities of the fellowship are as follows: - Course of lectures on diagnostics and treatment of acute radiation injuries - Everyday clinical examination of patients in the department of acute radiation disease or in the department of local radiation injuries - Training in applying and understanding the results of methods for estimation of radiation doses (cytogenetic, ESR, whole body counter) During the last three years Burnasyan Federal Medical Biophysical Center provided fellowships in the area of diagnosis and treatment of radiation injuries for 2 physicians from Georgia, 1 – from Byelorussia. This experience shows that fellowship at the base of Burnasyan Federal Medical Biophysical Center for specialists from developing countries has some very important advantages: common thecnological base, Russian language, life-style.

Internal Disometry and Bioassay

P.21   USTUR Case 0102 CT Image Processing Techniques For Voxel Phantom Development G. Tabatadze*, Idaho State University - Health Physics ; R. Brey, Idaho State University - Health Physics; T. James, United States Transuranium and Uranium Registries, Richland WA; D. Theel, Portneuf Medical Center, Pocatello ID; S. Todd, Portneuf Medical Center, Pocatello ID

Abstract: A 3D voxel model of the United States Transuranium and Uranium Registries’ (USTUR) Case 0102 241Am phantom (first whole body donation case in 1979) is described. Half of the donated skeleton is encased in tissue equivalent plastic and serves as a unique “human phantom” to calibrate external counting systems at United States Department of Energy (USDOE) and other laboratories world-wide. The resolution of the original CT (Computed Tomography) images of the USTUR Case 0102 Phantom is high (0.5 millimeters) which provides the physical basis for defining precisely the internal structure of the bones. Dicom images of the phantom (head, torso, arm, and leg) have been segmented using the Eclipse® radiotherapy planning software. This has a powerful automatic segmentation feature. The three-step segmentation procedure involved: defining the regions of interest as well as CT numbers for different anatomical structures; auto-segmenting the Dicom images, and; checking manually and correcting any errors in the auto segmentation results. Each Dicom image was segmented into the following regions of interest: air pockets, cortical bone, bone cavities (marrow/tabecular spongiosa), and soft tissue subdivided into ‘light’ and ‘regular’ regions to represent inhomogeneities (artifacts) that occurred when the case 102 phantom was cast in nominal ICRU tissue equivalent plastic. The range of CT numbers in each region of interest was replaced by a single characteristic CT number. The 3D surface models for each phantom (Non-Uniformal Rational B-Spline, NURBS) were created with Rhinoceros® software. Finally, these images were voxelized using MATLAB® into virtual (computational) phantoms. The application of these virtual phantoms to simulate the experimental response of external planar germanium detectors is discussed.

P.22   Critical Evaluation of( Pu-239)O2 Wound and Lymph Node Retention Predicted by NCRP 156’s Recommended Biokinetic Transfer Rates N. Chelidze*, Idaho State University - Health Physics ; R. Brey, Idaho State University - Health Physics; T. James, United States Transuranium and Uranium Registries

Abstract: A Powerbasic (PBCC 4.03) code has been developed to implement explicitly the general structure of the NCRP Publication 156 recycling biokinetic wound model, which partitions accidentally injected material into four characteristic initial states: (1)fragment, (2)particles, aggregates & bound state, (3) colloid & intermediate state, and; (4) soluble. This was bench-marked (quality assured) against compartmental retention values calculated separately for each of the four possible material states for Pu-239 (as a function of time after intake) by several European institutions, The new code also implements simultaneously the ICRP Publication 67 systemic biokinetic model for plutonium, to calculate the daily excretion of plutonium in urine as a function of time resulting from the combined blood uptake kinetics (from the wound ) and that of plutonium subsequently transferred to body organs. The utility of the NCRP wound model structure (and recommended inter-compartmental transfer rates) for predicting the wound and axillary lymph node retention measured for USTUR Case 0262 was examined. This worker died 33 y after receiving an accidental finger-puncture wound contaminated with (Pu-239)O2 particles and other plutonium material. A previously published empirical analysis of the data available in this case yielded four distinguishable phases of wound clearance, varying in characteristic rate over 5 orders of magnitude. The ‘mechanistic’ analysis carried out here examines the hypothetical fractionation between material states represented in the NCRP 156 wound model that is needed to ‘fit’ the USTUR Case 0262 data and the goodness-of-fit so obtained.

P.23   Development Of Calibration Phantoms For Newborns And Small Children V Sinha*, Idaho State University ; JT Harris, Idaho State University

Abstract: The experiences from several radiological accidents over the last few decades (e. g. Chernobyl, 1986, and Goiania, 1985) have shown that not only occupationally radiation-exposed adults but also children of all age groups can be affected. In addition, children could be exposed during an explosion of a radioactive dispersal device during a terrorist attack. Whole-body counting facilities throughout the world are well prepared to measure older children and adults, but suffer in the case of infants and toddlers from the lack of options to optimize their calibrations. A series of calibration phantoms representative of adult persons and larger children is commercially available or can be constructed from easily accessible materials. However for newborns and children up to 20 kg, very few phantoms have been described which approximate the infant body. The aim of this work was to improve this dilemma. The two main results included the compilation of anthropomorphic data for children and the construction of a modular set of calibration phantoms for weights between 3 and 20 kg from easily accessible and inexpensive materials. The phantoms are made from commercially available freeze packs which can be filled with a radionuclide solution. By using calibrations based on more realistic phantoms, internal radioactive contaminations in children can be measured more reliably than with the standard calibrations used for adults.

P.24A   Inhalation of Highly Insoluble Pu: Case Studies from the Rocky Flats Pu Fire M. Avtandilashvili*, Idaho State University - Health Physics ; R. Brey, Idaho State University - Health Physics; T. James, United States Transuranium and Uranium Registries; A. Birchall, Health Protection Authority, United Kingdom

Abstract: The United States Transuranium and Uranium Registries (USTUR) includes several whole- and partial-body donations from workers involved in the Rocky Flats Plutonium fire in 1965. This fire resulted in Pu contamination over about 6,500 m2 of working area with airborne Pu concentrations ranging from 3.7×10-2 Bq/m3 to greater than 3.7×104 Bq/m3. About 400 workers were monitored for their potential exposure to highly insoluble “high fired” plutonium dioxide particles (with a measured mass median physical particle diameter (MMD) of 0.32 micrometer). Several of the employees had intakes exceeding the contemporary permissible lung burden of 592 Bq (16 nCi ) by a factor of 1 to 17. The USTUR’s follow-up of relatively highly exposed individuals over several decades indicates that the inhaled plutonium is retained in the lungs significantly longer than expected for insoluble “Type S” plutonium [as characterized by the International Commission on Radiological Protection’s (ICRP) Publication 66 Human Respiratory Tract Model (HRTM)]. This phenomenon has been referred to as “Type Super S” absorption behavior, although the mechanism for this very long particle retention has not yet been established. Several cases with long follow-up and minimal influence of additional exposures were selected from the USTUR database. These were used to evaluate simultaneously the intake amounts, lung absorption rates and long-term particle transport (clearance) rates from the lungs, with their associated uncertainty distributions. The results are discussed.

P.24B   Measurement of Internal Exposure for Nuclear Medicine Workers involved in I-131 handling in Korea W.K. CHO*, Korea Institute of Nuclear Safety (KINS), Korea. ; K.J. LIM, Korea Radioisotope Association (KRA), Korea; K.H CHUNG, Korea Institute of Nuclear Safety (KINS), Korea.

Abstract: In Korea, safety management for internal exposures has been only applied to nuclear power plant workers. As a feasibility study for safety regulation of internal exposure of non-nuclear power plant workers, internal exposure measurements of nuclear medicine workers handling I-131 were performed. Considering the permitted amount of use and the dose coefficient, I-131 is expected to give the largest exposure to workers. Total of 306 measured data by thyroid uptake monitoring method were obtained for 76 workers and it was evaluated that about 8% of workers are certain to be exposed during the distribution work of I-131. More than 300 urine sample from the I-131 distribution workers has been measured and analyzed using NaI (Tl) scintillation detector. In urine sample analysis, it was evaluated that around 4.4% of workers are certain to be exposed by I-131. It is necessary to introduce regulatory measures to control the internal exposure of non-nuclear power plant workers, especially for the nuclear medicine workers handling unsealed radioisotopes such as I-131, Sr-89, Mo-99/Tc-99m etc.

P.24   An Updated Evaluation of data from the 1980 Statistical Analysis of Plutonium in U.S. Autopsy tissue. D.C Mecham*, Idaho State University ; R.R. Brey, Idaho State University; J.J. Shonka, Shonka Research Associates

Abstract: In a 1980 paper in the Health Physics Journal, T. Fox analyzed the results from tissues of over 900 individuals from various regions in the United States. The objective was to determine the level of Plutonium-239 in these tissues due to global fall-out from weapons testing. A comparison was made between 7 regions, including non-worker residents in the vicinity of Los Alamos National Laboratory (LANL). It was concluded that: “…no geographic difference in any of the tissue concentrations of plutonium was observed amongst the regions.” This study discusses a re-evaluation of the data from the Fox study. A regrouping of the residents living near LANL into pre and post 1960 sample groups, a reassessment of excluded outliers and an evaluation of liver to bone ratios of plutonium(Pu) is evaluated to examine the hypothesis: Higher concentrations of Pu-239 are present in residents living near LANL in the 1948-60 time frame in comparison to those in the post-1960 time frame and residents living in the 7 other regions.


P.25   BREMSSTRAHLUNG EXPOSURE FOR DNA, RNA AND RETINA HC Manjunatha bharadwai*, Bangalore university

Abstract: The Bremsstrahlung exposure for DNA, RNA and retina from patient administrated by beta nuclide has been calculated by extending the national council for radiation protection. We have estimated the specific Bremsstrahlung constant ( ), Probability of energy loss by beta during Bremsstrahlung emission (PBr), Bremsstrahlung activity (ABr) and Bremsstrahlung yield (IBr) for various energy. ABr is the activity above which patient should remain hospitalized on the basis of projected Bremsstrahlung dose. IBr for DNA, RNA and retina is estimated using tabulated results of IBr given for various elements at various energies by Lucien Pages [15]. This data may be useful in the analysis of Bremsstrahlung dose where beta emitting nuclide is involved in medical therapy.

P.26   Site-specific skeleton voxel model representing Chinese Reference adult Man and its absorbed dose for idealized photon exposures LIU Liye*, Tsinghua University ; ZENG Zhi, Key Laboratory of Particle & Radiation Imaging (Tsinghua University), Ministry of Education; LI Junli, Key Laboratory of High Energy Radiation Imaging Fundamental Science for National Defense; ZHANG Binquan, China Insitute for Radiation Protection; QIU Rui, Tsinghua University

Abstract: As part of an effort to develop a Chinese Reference adult Man voxel model (CRM), site-specific skeleton voxel model is built upon previous individual model (CNMAN). The whole skeleton is divided into 19 site-specific bones and bone groups as described in ICRP 89 publication, such as cranium, mandible, cervical spine, thoracic spine, lumbar spine, and so on. Each bone is then sub-segmented into cortical bone and spongious. The total skeleton mass is adjusted according to Asian reference value. Various bone-site-specific tissue masses, including red bone marrow, yellow bone marrow, cortical bone and trabecular bone, are derived using distribution data from ICRP 70 and 89 publications. Absorbed dose coefficients of active bone marrow of the resulting site-specific skeleton model are calculated for six idealized photon exposures using Monte Carlo transport code MCNP. The calculated values are a little higher than data of ICRP reference voxel model (Zankl, 2007) and the data of ICRP 74 publication (stylized model). It shows that dose coefficients of CRM are closer to the values of ICRP reference voxel model than to those of ICRP 74. It is also found that active marrow dose coefficient for AP exposure in ICRP 74 publication is believed to be underestimated compared with other exposure conditions.

P.27   Organ Dose Estimation for Computed Tomography Examinations Kwang Pyo Kim*, Kyung Hee University

Abstract: The number of Computed Tomography (CT) scans performed in the US has increased more than 300% in just over a decade. In addition, radiation dose from CT examination is much higher than conventional x-ray radiography (hundreds of times). CT scans provide great medical benefits. However, they result in radiation exposure to patients and potential cancer risk. Therefore, it is necessary to estimate cancer risk from CT scans. Organ dose rather than effective dose is needed for the risk estimation. However, most studies have provided only effective dose estimation. The objective of this study is to estimate radiation dose to specific organs from different CT scans in the US. Organ doses were estimated using the mean technical parameter settings (e.g. current-time product, peak potential, scan length) used in the US for each scan type. The parameters were taken from a nationwide survey (NEXT 2000 survey). Such parameters were not available in the NEXT survey for some CT scan types, including coronary artery calcification, virtual colonograhy, and CT angiography (CTA). For these, protocols in the recent literature were reviewed. Sex specific organ doses were estimated with these parameters using CT-expo for six of the most commonly used scanners. Radiation doses are high for organs entirely exposed to direct beam (~ > 10 mGy), moderate to organs partially exposed to directly beam or organs at vicinity to the direct beam (~ 1 - 10 mGy), and very small to the organs away from the scan region (< 1 mGy). Organs/tissues distributed across whole body (e.g., skin, bone surface, bone marrow) also received relatively high radiation doses. Organ specific doses are 44 mGy to brain for head scan, 20 mGy to lungs for chest scan, 20 mGy to stomach for abdomen scan, 18 mGy to bladder for pelvis scan, and 78 mGy to thyroid for cervical-spine scan, and 55 mSv to lungs for coronary CTA.

P.28   Minimization of radiation dose to operators performing cardiac catheterization procedures Kwang Pyo Kim*, Kyung Hee University ; Donald Miller, Uniformed Services University

Abstract: Cardiac catheterization procedures using fluoroscopy result in high scattered radiation dose to the cardiologist. Much has been written about how to properly use fluoroscopy equipment during cardiac interventions in order to optimize patient radiation dose. However, less attention has been paid to factors that affect operator radiation dose, and how to minimize operator dose. We review what is known about radiation exposure to physicians who perform cardiac interventions and discuss various factors that affect their exposure. There are wide variations in radiation dose up to 1000-fold per procedure (effective doses ranging from 0.02 - 38.0 uSv per procedure for diagnostic catheterization). Despite extensive improvements in equipment and technology, there has been little or no reduction in dose over time. The wide variation and lack of reduction in operator doses strongly suggests that more attention must be paid to factors influencing operator dose. Numerous patient, physician, and shielding factors influence operator dose to different degrees (changing operator dose from less than 100% - up to an order of magnitude). Principal among these are the fluoroscopy system and its operation. Some factors, such as patient and lesion characteristics and operator height and gender, are not under the operator's control. But many are under the control of the operator. Operators can change some of these factors immediately, at minimal or no cost, with a substantial reduction in dose and potential cancer risk. These include procedure technique, catheter choice, catheter insertion site, operator positioning, and the appropriate use of personal protective devices. Awareness of one's own radiation dose is essential, as it provides motivation for incorporating changes in one's practice that will result in a lower radiation dose.

P.29   Fluoroscopic Event Notification: An Automated Follow-up System JL Miller*, Mayo Clinic ; GM Sturchio, Mayo Clinic; KA Fetterly, Mayo Clinic; BA Schueler, Mayo Clinic

Abstract: The State of Minnesota promulgated a radiation protection regulation requiring that the registrant have a “procedure to ensure appropriate potential skin injury and follow-up information is given to the patient” in cases where the patient’s skin entrance exposure dose exceeded 6,000 milligray (600 rad). In addition, it requires that these cases be reviewed by the facility’s Radiation Safety Committee. Radiation Safety staff in collaboration with medical physicists from Interventional Radiology and Cardiovascular Imaging developed a web-based application for the reporting and review of these “fluoroscopic event” cases. This poster outlines the process flow from initial entry and submission of the event by procedure room staff to Radiation Safety’s reporting of events to the Radiation Safety Committee. The data entry screens used for submitting information are presented. Examples of automatically generated e-mail messages to the principal physician, staff designated for follow-up, radiation safety staff, and the appropriate medical physicists are provided. The implementation of this system has streamlined the process; thereby, reducing clinical staff and radiation staff time in ensuring the proper and timely reporting of the events to the Radiation Safety Committee.

P.30   Early Medical Consequences of Radiation Incidents in the Former USSR Territory L. Ilyin, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; V. Soloviev, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; K. Kotenko, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; A. Bushmanov*, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency

Abstract: In the FMBC n.a. Burnasyan of the FMBA of Russia (former Institute of Biophysics, Moscow) within 20 years the work on radiation accidents analysis at the former USSR territory which caused human exposure with clinically significant consequences, has been performed. Since 1949 to 2007, at the former USSR territory, at least 355 radiation incidents followed by the clinically significant human exposure had happened, which resulted in clinically significant health effects in 757 victims. That exposed cohort included 350 acute radiation sickness (ARS) patients. From that cohort 100 patients had the diagnosis ARS of severity and high severity degree and in 196 ARS was aggravated by local radiation injuries (LRI). 407 victims were affected by local radiation injuries only. Totally, 71 radiation induced fatalities were observed within first 3-4 months after the irradiation. The analysis concludes that in the nuclear industry 229 radiation accidents occurred from the number of incidents mentioned with 447 persons exposed (45 fatalities). The data with the case histories of radiation incidents exposed persons was input into a digital data base, which serves as a decision making support tool in acute radiation effects diagnosing and treatment. The paper sums up the information on radiation incidents at the former USSR territory, which incidents are related to the human exposure of clinically significant effects. L.A.Ilyin, V.Yu.Soloviev, K.V.Kotenko, A.Yu.Bushmanov

P.31   Small doses of external irradiation and risk of brain vascular illnesses F. Torubarov*, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; N. Isaeva, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; Z. Zvereva, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; G. Dmitrieva, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency

Abstract: The problem about the influence of small doses on the central nervous system (CNS) functional state is widely discussing. Some researchers think that single or chronic ionizing radiation exposure in doses up to 1 Gr., is mere supraliminal and can not lead to the functional and structural changes in CNS. The other authors concede that at chronic irradiation, which is related to the discharge in case of dose accumulation in the range from 0,5 to 0.75 Gr. , functional deviations in CNS are possible. But these deviations shouldn’t transcend physiological norms and lead to working ability disturbance. Among the great number of potential harmful for the people health professional effects, the radiative factor has a special place. It is known for a long time about the risks of large radiative doses. But the development of atomic energetics, space researches, nuclear-waste disposal and especially Chernobyl NPP accident arouses interest to the problem of small doses radiation influence. Of late years Russian and foreign science gives consideration to the studying of small and minute radiative doses influence on humans. The health state value of those, who expose to the ionizing emission influence, is holding, and the specific features of their work are studying. In one of Russian works, which is devoted to the risking factors assessment of brain vascular illnesses development among the workers in radiochemical production, high – risk of the development of the illnesses among the workers from the examined factory is conditioned by the complex of causes. Among them, as the issue showed, one of the leading positions occupies high psychoemotional working intensity, connected with particular working conditions in radiochemical production. There are some literary facts that one of the main psychogenic factor among the nuclear industry production staff is environmental radiation hazard awareness. At that, there are no any single opinion about the nervous system pathology origin because of the long influence of ionizing radiation in small doses. We have examined 1996 ( 1771 male and 225 female) workers from different Russian nuclear heating plants. According to the occupational aspect, the examined contingent was presented as by operating as by technical personnel. The age of the examined was from 35 to 71 years. All of them were exposed to occupational influence of ionizing radiation in small doses. The total external gamma irradiation cumulative dose varied from 0.08 to 1177.9 mSv . In our work the system of bone vacular abnormalities prediction, and particularly of ischemic stroke is used. This system was formed during the Fremingham research ( Massachusetts, USA) and we adapted it to solve the posed – problems. It is based on prognostic tables, developed for men and women of different age-groups, from 35 years old. These groups include 5 most informative risk factors of brain vascular illnesses development: Systolic, arterial, hypertension ( systolic blood pressure higher than 140 mm Hg); Hypercholesterolemia (cholesterol concentration in blood more than 6.98 millimole/ l); Hyperglycemia ( glucose concentration in blood more than 6.1 millimole/ l); Electrocardiographic signs of left ventricle of heart hypertrophy; smoking ( it is a risk factor and it doesn’t depend on smoked cigarettes amount). High risk of brain vascular pathology is detected at 899 workers ( 820 men and 79 women) and it is 45 %. 457 (50.8%) men from them, do not have any illnesses that lead to brain vascular pathology and some of them are almost healthy. To evaluate the radiation factor contribution to the brain vascular pathology development a correlated analysis between brain vascular illnesses development on the one hand, and the total external gamma irradiation cumulative dose, on the other hand. In the course of the research, among the workers of different Russian nuclear heating plants, we didn’t reveal the connection between direct effect of ionizing radiation and brain vascular illnesses origin. The data, we got in the research issue indicate of cerebrovascular illnesses risk increases by age, and the greatest number of people having risk more then population one, are age-groups of 45-49 years and 50-54 years, both men and women, that is the most capable age. The dominating risk factors in our research are: arterial hypertension and smoking, regardless of age and sex. F.S. Torubarov, N.A. Isaeva, Z.F. Zvereva, G.E. Dmitrieva

P.32   ABNORMAL HEAD PENETRATING IRRADIATION BY HIGH ENERGY PROTON BEAM F. Torubarov*, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; Z. Zvereva, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; N. Isaeva, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency ; G. Dmitrieva, Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency

Abstract: The research worker of the Institute of physics of high energy, a young, healthy man of 36 years, as a result of safety engineering breakdown in June 1978, run the short-term high energy proton beam impulse action. The size of high energy proton beam was 2x3 mm.,the ionizing radiation entry dose was 2000 Gr. Vital centers and brain structures were not touched, the man survived. The proton beam came into the back zone of the head, to the left of sagittal line. It damaged occipital bone, passed through the left end lobe, mediobasal parts of the left temporal lobe, upper parts of the left periotic bone, damaged middle ear, Fallopian aqueduct, the left zygomatic bone, entered in upper angle of the left nasolabial fold. The damaged head tissues in the proton beam entry and outlet area healed rather quickly – in 3 months. But in the left periotic bone and in middle ear (the structures located to the beam motion direction) inflammatory - necrotic process had begun. It lasted during the whole 20-year observation period. Low - intensity inflammatory -necrotic process of the middle ear and adjacent parts of the left periotic bone got the source of a protracted stimulation in adjacent parts of brain- left temporal lobe. In time it caused posttraumatic epilepsy formation. Constant therapy didn’t remove the manifestation of the illness, as the inflammatory -necrotic process in the left temporal lobe continued. At the age of 51 duodenum stomach ulcer and arterial hypertension appeared, with the lapse of time their clinical presentations got stronger. In spite of a severe disease, the patient managed to keep up his psycho physiological and social adaptation level during a long time – 20 years. According to the psycho physiological examination, which was carried out in 10 years, the psychophysiological adaptation level was normal. In 20 years after the injury, the psycho physiological adaptation level corresponded to the lower norm limit. The patient worked as a senior researcher in the Institute of physics of high energy, was married and had a grown-up son. Torubarov F.S., Zvereva Z.F., Isaeva N.A., Dmitrieva G.E.

P.33   Photochemical Delivery of Bleomycin in Malignant Glioma Cells JW Blickenstaff*, UNLV ; V Vo, UNLV; H Hirschberg, UC, Irvine; SJ Madsen, UNLV

Abstract: Photochemical internalization (PCI) is a new technology in which photosensitizing drugs are used to improve the utilization of macromolecules in cancer therapy in a site-specific manner. The concept is to localize the sensitizer and the macromolecule (e.g. chemotherapeutic agents such as bleomycin) in endocytic vesicles of target cells (e.g. infiltrating glioma cells) and then excite the photosensitizer (e.g. aluminum phthalocyanine disulfonate; AlPcS2a) with light. This releases the endocytically encapsulated macromolecules into the cytosol, but only in the irradiated area. The rapid attenuation of light in the brain should lead to minimal side effects since the effect is localized to the irradiated area. The utility of PCI for the treatment of malignant gliomas was investigated in vitro using both an F98 rat and ACBT human glioma cell line. The cytotoxicity of the AlPcS2a – based PCI of bleomycin was compared to: (1) AlPcS2a–PDT (18 h incubation time), and (2) bleomycin (4 h incubation time). In the PCI studies, monolayers were incubated in AlPcS2a for 18 hours followed by 4 hour bleomycin exposure. In both PCI and PDT studies, cells were exposed to 670 nm light. Toxicity was evaluated using colony formation assays as well as spheroid growth analysis. F98 rat glioma cells in monolayer were found to be susceptible to the effects of both AlPcS2a-PDT and bleomycin. AlPcS2a-PDT was found to be particularly effective when light was delivered at a low irradiance of 5 mW cm-2. In this case, a radiant exposure of 6 J cm-2 resulted in only 3% survival. Bleomycin was found to be toxic at relatively low concentrations – incubation of F98 cells in 5 µg ml-1 for 4 hours resulted in 5% survival. A total radiant exposure of approximately 2 J cm-2 (50% survival) was found to be optimal for the PCI effect in this experimental system. ACBT spheroid experiments are being conducted to determine if similar effects are seen in a more complex in vitro model. PCI has shown promising results in both monolayer and spheroid models.

Operational Health Physics

P.34   Use of a Database for Accurate Shipment Labeling and Generation of Shipment Forms Jamie Miller*, Mayo Clinic ; Rodney Landsworth, Mayo Clinic; Kelly Classic, Mayo Clinic

Abstract: Department of Transportation (DOT) regulations are extremely detailed and, for those not routinely shipping packages, can be quite onerous. Additionally, routine DOT audits for our internal shipping program (over-the-road from one Mayo Clinic Rochester facility to another Mayo Clinic Rochester facility) consistently showed minor paperwork errors. In an effort to overcome these two issues, we chose to develop a simple addition to our database that, with the entry of a few key pieces of information, would tell the shipper the type of label to use and would print out the shipping paper. The developed program allows for up to four packages to be entered (per shipping paper) with each package containing one or two radionuclides. The current radiation safety office database contained basic radionuclide information. We added Limited Quantity and Reportable Quantity values to allow the new program to determine the appropriate shipping description wording upon entry of the radionuclide, form, activity, and surface and one meter dose rates. For activity entry, the unit can be gigabequerel or millicurie; the computer will convert to proper units for the labeling. Dose rates are entered in milliroentgen (or millirem) per hour. Although it is a relatively simple, straight-forward program, the shipping portion of the database will help ensure accuracy in shipment paperwork whether completed by a novice or a DOT expert.

P.35   Photon Response of Savannah River Site Instrumentation from 38 to 1300 keV D. A. Wagoner*, Savannah River Site

Abstract: Personnel at the Department of Energy Savannah River Site rely on many different types of portable radiation detection instrumentation to complete work assignments within prescribed dose limits. Generally radiation detection instruments are calibrated using Cesium-137, because of its moderately energetic gamma ray at 662 keV. However, it is very common for an instrument’s response to fluctuate greatly with respect to photon energy, especially below 300 keV. In this study, the relative response of the following instruments was tested; Eberline RO-20, RO-2S-1, RO-7, Teletector 6112B, Bicron MicroRem LE, and the Mk2 EPD, over a photon range of 38-1300 keV. In order to achieve this range of photon energies, the following irradiator systems were used; an X-ray Irradiator calibrated using National Institute of Standards and Technology (NIST) techniques, an Americium-241 Irradiator, and a Low Scatter Irradiator equipped with Cobalt-60 and Cesium-137 sources. To determine relative response, a small population of each type of instrument was irradiated with photons of various energies in an assortment of orientations. The instrument’s response, at each photon energy in each orientation, were averaged and normalized to the photon energy and orientation in which that instrument is calibrated. From these results, a plot of relative response vs. effective energy was constructed for each instrument. These graphs illustrate the importance of instrument orientation and, in some cases, significant variation of instrument response at low energies. In addition to low-energy photon dependence of radiation instrumentation, the exposure to dose equivalent conversion factor must be taken into consideration. For moderately energetic photons, the exposure to dose equivalent conversion factor is essentially 1. However, for low-energy photons this conversion factor can exceed 1.6. In a situation where the source term consists of moderately energetic photons, it is acceptable to use an exposure rate instrument to set dose rates. However, if the source term is primarily low-energy photons, an instrument which measures dose equivalent rate must be used.

P.36   Dose and Dose Equivalent Rate Calculations from a Solar Energetic Particle Event using Earth-Moon-Mars Radiation Environment Module (EMMREM) M PourArsalan*, University of Tennessee, Nuclear Engineering Department ; L. W. Towsend, University of Tennessee, Nuclear Engineering Department; N. A. Schwadron, Boston University, Astronomy Dept; K Kozarev, Boston University, Astronomy Dept; M Al-Dayeh, Southwest Research Institute

Abstract: The central objective of the EMMREM (Earth-Moon-Mars Radiation Environment Module) is to develop a numerical model for completely characterizing the time-dependent radiation environment in the Earth-Moon-Mars and Interplanetary space environments. The Module includes a 3D energetic particle transport model (EPREM), and utilizes a version of the space radiation transport code (BRYNTRN) developed at NASA Langley Research Center. With the initial setup of the EMMREM (Earth-Moon-Mars Radiation Environment Module) framework in place, we are turning to performing realistic simulations with observations from Solar Energetic Particle events for module testing and as an example of the module capabilities. In this work we present and discuss the EMMREM (Earth-Moon-Mars Radiation Environment Module) predictions for the dose rate, dose equivalent rate and accumulated dose in space, throughout an event, for various aluminum and water shield thicknesses. These dose and dose equivalent data, when folded with the actual shielding distributions of a spacecraft or surface habitat, and folded with the distribution of overlying tissues providing body self-shielding for human organs, will permit time-dependent estimates of organ exposures for human crews in deep space.

P.37   A Review of Programmatic Upgrades Recently Implemented in the Oak Ridge National Laboratory Bioassay Monitoring Program D.A. McLaughlin*, Oak Ridge National Laboratory ; T.J. Gillespie, Oak Ridge National Laboratory; D.E. Perkins, Oak Ridge National Laboratory; J.R. Benson, Oak Ridge National Laboratory

Abstract: For over seven decades the Radiological Protection Program at the Oak Ridge National Laboratory (ORNL) has operated a Bioassay Monitoring Program. The conduct and operation of this program has evolved to satisfy changing regulatory requirements, incorporate technological advances, and apply contemporary management strategies. Spurred by a commitment for continuous improvement, revisions to 10 CFR 835, and a critical review by the U.S. Department of Energy Inspector General, the most recent evolution of the ORNL Bioassay Monitoring Program was implemented in January 2009. Operationally, the changes incorporate bioassay decision making into the Laboratory’s Integrated Safety Management System as part of pre-job hazard analysis. Most importantly, a process for performing prospective determinations of the likelihood that radiological work activities could deliver internal doses greater than the 10 CFR 835.402 bioassay monitoring threshold of 0.001 Sv (100 mrem) Committed Effective Dose Equivalent has been standardized. Prospective determinations are completed by radiological engineers using a Microsoft Excel-based software application which applies ANSI N13.39 methodologies to estimate exposure potential based on radiological source term information, applied engineering controls, and other job-specific parameters. As a further aid towards developing bioassay monitoring strategies, the prospective determination software identifies dominant dose contributors and compares nuclide-specific bioassay detection levels against nuclide retention and excretion patterns. A random confirmatory bioassay monitoring program, operated under the provisions of 10 CFR 835.401, was also initiated to verify the effectiveness of this approach. Early operating experience revealed several benefits including increased collaboration between dosimetry and radiological work planning staff, a decrease in the number of radiological work permits requiring bioassay monitoring, clarified dosimetry requirements, and improved worker understanding.

P.38   Numerical Solutions For Confidence Intervals When The Sample Is Counted An Integer Times Longer Than The Blank WE Potter*, Consultant, Sacramento ; J Strzelczyk, University of Colorado, Denver

Abstract: Past computer solutions for confidence intervals in paired counting are extended to the case where the ratio of the sample count time to the blank count time is taken to be an integer, IRR. Previously, confidence intervals have been named Neyman-Pearson confidence intervals; more correctly they should have been named Neyman confidence intervals or simply confidence intervals. The technique utilized mimics a technique used by Pearson and Hartley to tabulate confidence intervals for the expected value of the discrete Poisson and Binomial distributions. The blank count and the contribution of the sample to the gross count are assumed to be Poisson distributed. The expected value of the blank count, in the sample count time, is assumed known. The net count (OC) is taken to be the gross count minus the product of IRR with the blank count. The probability density function (PDF) for the net count can be determined in a straightforward manner. When OC is observed, two confidence intervals for the expected net count that can be constructed are [0.0, x] and [y, +infinity). In the former confidence interval x is that value for the expected net count that has probability beta1 of obtaining OC or less net counts in which case the confidence level is (1.0-beta1)100%. In the later confidence interval y is that value for the expected net count that has probability beta2 of obtaining OC or greater net counts in which case the confidence level is (1.0-beta2)100%. The quantities beta1 and beta2 are taken to be in [0.5, 0.999]. If y is > 0.0, [y, x] is a (1.0 - beta1 - beta2)100% confidence interval for the expected net count when OC is observed. The code checks that the sum of probabilities, both when there is no activity in the sample and when there is an amount equal to the computed confidence limit in the sample, equals 1.0. Also the code checks that the computed expected net count, when there is no activity in the sample, equals 0.0.

P.39   Classification of Radiation Devices for Industrial Application and Measurement of Radiation Dose in Accident Conditions W.K. CHO*, Korea Institute of Nuclear Safety (KINS), Korea ; K.S. SEO, Korea Institute of Nuclear Safety (KINS), Korea; B.C. KOO, Korea Institute of Nuclear Safety (KINS), Korea; C.B. KIM, Korea Institute of Nuclear Safety (KINS), Korea

Abstract: Radiation devices for industrial application were classified into Class I through Class IV considering the device performance, measured radiation dose at maximum performance, required safety mechanism, operation procedures and anticipated accident dose from accident scenario etc. Representative devices were selected for each class and the leakage radiation level in normal operation condition was measured using typical detection equipments. In an accident condition, radiation exposure was measured using TLD and measuring points were determined by simulating accident situations. For the X-ray Fluorescence (XRF), the representative device of Class I, the leakage radiation in normal operation was less than 0.2 ¥ìSv/hr and the measured radiation dose in accident condition was less than 0.55 mSv. Baggage inspection device, the representative device of Class II, can give the leakage radiation up to 10 ¥ìSv/h and the accident radiation dose to 3.41 mSv. Inspection device for various parts' defect, the representative device of Class III, the leakage radiation in normal operation was less than 0.2 ¥ìSv/h and the measured radiation dose in accident condition was less than 3.83 mSv. For Class IV, the representative device is linear electron accelerator of 9 MeV and the neutron dose was measured using BF3 moderating detector. The leakage neutron dose in normal operation was less than 2.5 ¥ìSv/h and the maximum accident dose measured with BF3 detector has reached to 80 mSv.

Power Reactor

P.40   Evaluation of Neutron Flux and Gamma Dose Rates at the Irradiation Cell of the Texas A&M Nuclear Science Center Reactor L Vasudevan*, Nuclear Science Center, Texas A&M University, College Station, TX ; J Newhouse, Nuclear Science Center, Texas A&M University, College Station, TX; J Remlinger, Nuclear Science Center, Texas A&M University, College Station, TX; W. D Reece, Nuclear Science Center, Texas A&M University, College Station, TX

Abstract: Nuclear Science Center (NSC) at the Texas A&M University houses a 1 MW TRIGA pool type reactor for use in research, teaching, as well as for the production of isotopes. The irradiation cell (dry cell) adjacent to the reactor pool is a shielded structure designed and used mainly for the irradiation of either small biological samples or other large objects with the reactor as the source and sometimes with a lanthanum (La-140) gamma source. A sealed irradiation window located on the West side wall separates the reactor pool and dry cell. The reactor can be moved within the water pool, position against the dry cell and perform steady state operation at 1 MW for irradiating samples with neutrons. The reactor can be placed in the shut down mode after several hours of operation to use as a sole gamma source also. For obtaining gamma radiation exposure on the order of a several kGy, a La-140 gamma source can be produced using the reactor and can be placed next to the window for gamma irradiation and this practice has been in use for past several years. The reactor core recently underwent a complete conversion from the FLIP (Fuel Life Improvement Program) HEU fuel to standard LEU fuel. It then became imperative to characterize the neutron flux at several experimental locations including the dry cell. One of the tasks attempted in this paper is to compute the neutron flux at locations inside the dry cell using MCNP-5 code. The code was used to model the reactor core against the dry cell window at full power of operation. The neutron flux obtained inside the cell which when compared with published values in the literature for the past FLIP fuel showed a variation of less than 10%. Measurements were also made with a set of activation foils and the activities agree with the MCNP computed results. Experimental determination of gamma dose rates and the corresponding integrated dose was made using an ion chamber in the dry cell and with the reactor positioned against the window after 7 hrs of operation. Moreover, a La-140 source was made by irradiating approximately 2000 grams of lanthanum oxide in the reactor for about 10 hrs, then positioned against the cell and the integrated gamma dose was obtained using the ion chamber. A set of radio chromic films were placed at different locations on the window and gamma dose mapping with the La-140 source was also performed.

Waste Management

P.41   Long-Term Performance Of Transuranic Waste Inadvertently Disposed In A Shallow Land Burial Trench At The Nevada Test Site Gregory Shott*, National Security Technologies, LLC ; Vefa Yucel, National Security Technologies, LLC

Abstract: In 1986, 21 m3 of transuranic (TRU) waste was inadvertently disposed in a shallow land burial trench at the Area 5 Radioactive Waste Management Site on the Nevada Test Site. U.S. Department of Energy (DOE) TRU waste must be disposed in accordance with Title 40, Code of Federal Regulations (CFR), Part 191, Environmental Radiation Protection Standard for Management and Disposal of Spent Nuclear Fuel, High-Level, and Transuranic Radioactive Wastes. The Waste Isolation Pilot Plant is the only facility meeting these requirements. The National Research Council, however, has found that exhumation of buried TRU waste for disposal in a deep geologic repository may not be warranted when the effort, exposures, and expense of retrieval are not commensurate with the risk reduction achieved. The long-term risks of leaving the TRU waste in-place are evaluated in two probabilistic performance assessments. A composite analysis, assessing the dose from all disposed waste and interacting sources of residual contamination, estimates an annual total effective dose equivalent (TEDE) of 0.01 mSv, or 3 percent of the dose constraint. A 40 CFR 191 performance assessment also indicates there is reasonable assurance of meeting all requirements. The 40 CFR 191.15 annual mean TEDE for a member of the public is estimated to reach a maximum of 0.055 mSv at 10,000 years, or approximately 37 percent of the 0.15 mSv individual protection requirement. In both assessments greater than 99 percent of the dose is from co-disposed low-level waste. The simulated probability of the 40 CFR 191.13 cumulative release exceeding 1 and 10 times the release limit is estimated to be 0.0093 and less than 0.0001, respectively. Site characterization data and hydrologic process modeling support a conclusion of no groundwater pathway within 10,000 years. Monte Carlo uncertainty analysis indicates that there is reasonable assurance of meeting all regulatory requirements. Sensitivity analysis indicates that the results are insensitive to TRU waste-related parameters. Limited quantities of TRU waste in a shallow land burial trench can meet DOE performance objectives for disposal of TRU waste and contribute negligibly to disposal site risk. Leaving limited quantities of buried TRU waste in-place may be preferred over retrieval for disposal in a deep geologic repository.

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