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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Modeling Analysis for Nuclear Material Transport Package

This paper describes the computational models for the thermal analyses for 9975 nuclear material package under the Normal Conditions of Transport (NCT) and the Hypothetical Accident Conditions (HAC). The package 9975 is a DOE certified package that has been used in the DOE complex for transportation and long term storage of radioactive material. The NCT and HAC thermal analyses for the 9975 package were performed here by taking a computational heat transfer method. The HAC thermal models described here are identical to the models in the current DOE certified 9975 package. Thermal models in this work are for an undamaged package to meet all the regulatory requirements during an HAC thermal event. The model was created using modified shielding where the lead was encapsulated in a 0.036 inches stainless steel casing. The impact of this casing was studied earlier when the modification was proposed and it was found to have minimal impact on the component temperatures. Therefore, the analyses in this paper are good for packages with or without the shielding modifications. Analytical studies show that the realistic value for the solar absorptivity for the stainless steel surface fabricated from the rolling operation is about 0.5. The NCT analyses documented in this report use an absorptivity value of 0.5 to get more realistic component temperature results. Sensitivity studies were also performed to assess the impact of variability in the solar absorptivity for drum surface in the previous work for the NRC certification. The analysis in this paper is based on PuO{sub 2} in a food can. The food can is considered uncredited convenience containers inside the Primary Containment Vessel (PCV) of the 9975 package. PCV is located inside Secondary Containment Vessel (SCV). PuO{sub 2} has a density of 2 g/cc and a heat generation rate of 19 watts. This calculation evaluates the thermal performance for 9975 package with food can configuration during NCT and HAC per NRC regulations in 10 CFR Part 71. The objective of the calculation is to demonstrate that the maximum component temperatures are below their design limits. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Distribution Development for Residual Inventory at the New York West Valley Site - 20400

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351 ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into Agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States. NFS, a private company, built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations ended in 1972 and never reopened, leaving behind radioactive and chemical wastes. Operations led to contamination in a number of facilities and locations. Some of that contamination has migrated from waste disposal zones to other layers, formations, and features on and off the WNYNSC. Phase I decommissioning activities are ongoing and involve the removal of a number of areas and structures that have been associated with contamination. The purpose of this work is to outline the approach for characterizing contamination not associated with disposed wastes, contaminated structures, or specific releases. In this work, the term, residual radiological activity, is used to describe environmental contamination that exists subsequent to the completion of Phase I decommissioning activities, that is not associated with disposed wastes, contaminated structures, or specific releases. Contamination from the Site was quantified relative to data that characterize the concentrations of radionuclides that exist in background. Background concentrations are those present in the area but having no influence from Site related activities. The existence of residual radiological activity that is elevated relative to background has the potential to contribute to future risks to human health and the environment. As a consequence, the residual inventory information is used to inform the West Valley Probabilistic Performance Assessment (PPA) model to characterize potential future risks to human health and the environment. The centralized West Valley Data Management System (DMS) was the source of information for the data assembled in this analysis. The DMS is a fairly large compilation consisting of thousands of records from investigation studies, with sample dates ranging from 1990 to present. Samples from monitoring wells, boreholes, geoprobe studies, surface water, surface soils, storm water outfalls, ventilation stack filters, plant and animal tissues, and more are included in the DMS. Results are typically reported in units of activity per unit volume. For the purpose of the analyses presented here, all results were converted into consistent units of pCi per unit volume. Since 1990, data have been collected from various locations across the WNYNSC at different times with varying frequency over the course of several decades. As a consequence, a number of potential issues can arise with respect to the assembly of a dataset that is deemed adequate for the characterization of residual radiological activity. These issues were assessed and resolved to the extent possible through careful consideration of the properties of the distributions. The intent was to use data which characterize the current state of the Site. Radionuclides can be designated to one of several groups depending on their origin. In this work the groups considered were 1) Naturally Occurring Radioactive Material (NORM), 2) fallout, and 3) Other (including power plant, medical research, etc). This grouping is a useful construct with respect to the interpretation of fixed laboratory results. For example, NORM radionuclides that exist within a decay chain should have approximately equivalent distributions of concentrations if they are representative of background conditions. Insights such as these can be used as a check to identify sample results that need to be further investigated or omitted due to issues associated with reported results from fixed laboratory analyses. This type of analysis provided a foundation for the assessment of the adequacy of sample results for use in subsequent components of an assessment. The general process for the assessment of residual radiological contamination at the Site consists of a sequence of several steps. First, for each analyte, several statistical tests were performed to assess the weight of evidence against the null hypothesis that the mean of the distribution of concentrations was equal to zero. If the mean of the distribution of concentrations for a given radionuclide was not found to be greater than zero, then it was removed from consideration as a component of the residual radiological contamination. If there was significant evidence to reject the hypothesis of the mean being equal to zero, the second step was to compare the distribution of the data from the Site to that of the corresponding background. A suite of tests was used to compare the distributions of the site and background data. The results of these tests were collectively used to determine if site data are elevated relative to background. The third step was to develop distributions using a Bayesian framework to characterize the distribution of mean of the increment present above background for each of the radionuclides. The Bayesian model implemented allowed for the comparison of site-specific records to background concentrations to better approximate contamination attributed to the Site. A final screening step was employed for radionuclides that exceed background. This screening step compared 95% upper confidence limits (UCLs) from the increment distribution developed in the previous step to the risk screening levels. This approach yields a list of analytes that were determined to be elevated relative to background.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A Retrospective Assessment of Correlation between Occupational and Public Radiation Protection Programs Performance Metrics at a FUSRAP Site - 17546

A retrospective assessment of the occupational and environmental radiation protection programs includes the consideration of performance metrics at the qualitative and quantitative level. When key radiation protection (RP) program metrics are tracked over the course of several years, correlations may present themselves. Strong correlations between variables may be considered when establishing or validating baseline RP Program elements (i.e., controls, monitoring regimes, and additional ALARA measures) for the safe excavation and handling of naturally occurring radioactive materials comingled in soil-like wastes at FUSRAP sites. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Overview of the US Nuclear Science User Facilities

Specialized and often unique and expensive instrumentation, adapted for use on radioactive materials, is essential to conduct research in the area of nuclear energy. Much of this equipment is located in facilities dedicated for such work that have been built and maintained over many decades. The Nuclear Science User Facilities (NSUF) was established in 2007 under the US Department of Energy Office of Nuclear Energy (DOE-NE) in order to bring these capabilities together with those innovative ideas from researchers at universities, national laboratories, small business, and industry that might not have regular access to reactors, hot cells, transportation casks, radiologically qualified instrumentation, and associated expertise. Early descriptions of the originally named Advanced Test Reactor National Scientific User Facility have been given. Through a competitively reviewed proposal process, the NSUF offers access to these capabilities and expertise at no cost to the user to promote research that will address the mission of DOE-NE. As a national organization comprised of a number of partner facilities across the US and located at National Laboratory, university, and industry sites, the NSUF supports materials science studies directed at irradiation effects in nuclear fuels and materials as well as in-pile and post irradiation examination (PIE) instrumentation development. The NSUF has created a Nuclear Fuels and Materials Library (NFML) of previously irradiated materials that users may access for their studies. Expanding the library for future applications of advanced technologies and novel concepts is an important objective of the NSUF. The NSUF has established and maintains the Nuclear Energy Infrastructure Database (NEID) as a tool for both DOE-NE and the nuclear energy research community. The extensive catalogue of capabilities is intended to aid DOE-NE, NSUF, and researchers more effectively and efficiently utilize the available capabilities. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2011

Tritium plasma experiment: Parameters and potentials for fusion plasma-wall interaction studies

The tritium plasma experiment (TPE) is a unique facility devoted to experiments on the behavior of deuterium/tritium in toxic (e.g., beryllium) and radioactive materials for fusion plasma-wall interaction studies. A Langmuir probe was added to the system to characterize the plasma conditions in TPE. With this new diagnostic, we found the achievable electron temperature ranged from 5.0 to 10.0 eV, the electron density varied from 5.0 x 10{sup 16} to 2.5 x 10{sup 18} m{sup -3}, and the ion flux density varied between 5.0 x 10{sup 20} to 2.5 x 10{sup 22} m{sup -2} s{sup -1} along the centerline of the plasma. A comparison of these plasma parameters with the conditions expected for the plasma facing components (PFCs) in ITER shows that TPE is capable of achieving most ({approx}800 m{sup 2} of 850 m{sup 2} total PFCs area) of the expected ion flux density and electron density conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Traces of natural radionuclides in animal food

Naturally occurring radioactive materials are present everywhere, e.g., in soil, air, housing materials, food, etc. Therefore, human beings and animals receive internal exposure from radioactive elements inside their bodies through breathing and alimentation. Gamma radiation has enough energy to remove an electron from the atom and compromise the rearrangement of electrons in the search for a more stable configuration which can disturb molecule chemical bonding. Food ingestion is one of the most common forms of radioisotopes absorption. The goal of this work is the measurement of natural gamma radiation rates from natural radioisotopes present in animal food. To determine the concentration of natural radionuclides present in animal food gamma-ray spectrometry was applied. We have prepared animal food samples for poultry, fish, dogs, cats and cattle. The two highest total ingestion effective doses observed refers to a sample of mineral salt cattle, 95.3(15) μSv/year, rabbit chow, with a value of 48(5) μSv/year, and cattle mineral salt, with a value of 69(7) μSv/year, while the annual total dose value from terrestrial intake radionuclide is of the order of 290 μSv/year.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2019

Comparison of {sup 238}U, {sup 210}Po and {sup 210}Pb Speciation in Six Different Types of Soils

Speciation of {sup 238}U, {sup 210}Po and {sup 210}Pb in six different types of Syrian soil (Entisols, Inceptisol, Aridisol, Mollisols, Vertisols and Rock outcrops) was studied using a four-step sequential extraction procedure. The binding nature of {sup 238}U, {sup 210}Po and {sup 210}Pb was correlated with soil properties including soil particles, pH, EC, OM, sand, silt and clay. The results showed that the exchangeable and carbonate fractions of {sup 238}U did not exceed 10%. In addition, the {sup 238}U, {sup 210}Po and {sup 210}Pb fractions bonded to organic matter increased as the soil clay content decreased. The highest {sup 210}Po fraction was found to be bonded to soil Fe/Al/Mn-oxides for Entisols soils (56%) and silica for Aridisol soils (53%). Soil clay content showed a linear correlation with {sup 210}Po concentrations in soils ranging from 0.34 to 0.91. The results are considered useful in assessing the risks resulting from soils contaminated by naturally occurring radioactive material from the phosphate and oil industry.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Crush Strengths of Stainless Steel Honeycomb for Packaging

Honeycomb structures are used as energy absorbers to dissipate kinetic energy. Their low density to high axial strength and inexpensive cost make them ideal for the energy absorption application. Mechanical properties of lightweight materials are well known compared to more dense materials like stainless steel. Determines the viability of using 304 Stainless Steel honeycomb in future SRNL radioactive material packages. Using a combination of the material properties and testing procedures both static and dynamic crush strength of stainless steel honeycomb will be determined. Both static and dynamic crush strengths of stainless steel honeycomb need to be quantified because strain rate plays a role in the material strength. The stainless steel honeycomb being tested will be cut into sections of about 2.0 inches in width, 2.3 inches in length and 2.5 inches in height. The cell size of the material is 0.25 in and the ribbon thickness 0.003 in. The material has a yield strength of 48 ksi and an ultimate strength of 103.2 ksi, with an elastic modulus of 29,300 ksi. Compression testing to determine the static crush strength will use an Instron compression test machine. This Instron machine uses a 20 ksi capacity load cell, while recording data at varying low strain rates. Dynamic impact testing will involve dropping a 43.8 lb impactor from a height around 5 feet onto the stainless steel honeycomb. The honeycomb structure will be on a rigid surface and the impactor will be dropped via a quick release hook. Dynamic crush strength is determined from the weight of the impactor, height of the drop and difference in height of the honeycomb after the drop.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

The influence of cooling rate on condensation of iron, aluminum, and uranium oxide nanoparticles

Fundamental observations of particle size distributions are needed to develop models that predict the fate and transport of radioactive materials in the atmosphere following a nuclear incident. The extent of material transport is influenced by the time scales of particle formation processes (e.g., condensation, coagulation). In this study, we investigated the influence of cooling time scales on size distributions of uranium, aluminum, and iron oxide particles that are synthesized separately under identical run conditions inside the controlled environment of an argon plasma flow reactor. Two distinct temperature distributions are imposed along the flow reactor by varying the argon flow rate downstream of the plasma torch. The vaporized reactants of uranium, aluminum, and iron are cooled from about 5000K to 1000K before they are collected on silicon wafers for ex situ scanning electron microscope analysis. The microscope images show that the sizes of the largest aluminum and iron oxide particles heavily depend on the cooling time scales, whereas significant size variation with cooling rate is not observed for uranium oxide particles. In addition, the size distribution of aluminum oxide particles exhibits the broadest range among all three metal oxides studied. We performed simulations of particle size distributions using a kinetic model that couples gas phase oxidation chemistry with particle formation processes, including nucleation, condensation, and coagulation. The model results demonstrate the strong sensitivity of particle size distribution to different cooling histories (i.e., temperature vs residence time) along the flow reactor. In conclusion, the kinetic model also helps identify directions for future research to improve the predictions.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Light Water Reactor Sustainability Program: Survey of Models for Concrete Degradation

Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have predictive tools to address concerns related to aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to review and document the main aging mechanisms of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

The status and prospective of environmental radiation monitoring stations in Saudi Arabia

The use of nuclear technology requires an environmental monitoring program to ensure the safety of the environment, and to protect people from the hazards of radioactive materials, and nuclear accidents. Nuclear accidents are unique, for they incur effects that surpass international frontiers, and can even have a long lasting impact on Earth. Such was the case of the Chernobyl accident in the Ukraine on April 6, 1986. For that purpose, international and national efforts come together to observe for any nuclear or radioactive accident. Many states, including Saudi Arabia which oversees the operation of the National Radiation, Environmental and Early Monitoring Stations, The Radiation Monitoring Stations(RMS’s) are currently scattered across 35 cities in the country,. These locations are evaluated based on various technological criteria such as border cities, cities of high population density, wind direction, etc. For new nuclear power plants hovering around, it is strongly recommended to increase the number of radiation monitoring stations to warn against any threat that may arise from a nuclear leak or accident and to improve the performance of the existing RMS’s. SARA (Spectroscopic Monitoring Station for air) should be implemented due to the high sensitivity to artificial radiation, automatic isotope identification, free of maintenance, and fully independent due to solar power supply (incl. battery backup) and wireless communication (GPRS)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Exploring Actinide Nanocrystal Growth towards Defining 5f Surface Chemistry (Final Technical Report)

When particles are very small, down to the nanometer length scale, they display unusual properties not typical of larger materials. Most of the atoms in these structures are at the surface of the particles, which give them different electronic properties. While these unusual properties have been studied in most of the periodic table, little is known about how the electronic properties of the actinides, such as neptunium and plutonium, behave when they become so small. This planned research will study growth pathways of actinide nanoscale particles. Interestingly, the shapes of the particles change depending on the specific actinide species, spanning from thorium through americium, despite having otherwise identical arrangements of atoms. This project will use differences in the shapes and growth pathways of different actinide particles to learn about trends in electronic properties at surfaces across the actinide row. Growth pathways will be determined using a combination of synthesis and the use of advanced X-ray characterization tools. These X-ray characterization tools can provide information about electronic properties and also local bonding characteristics. This project will also inform how to make unprecedented actinide oxide nanoparticles using very small quantities of material. This will enable safe working with radioactive materials and the study of materials that never before have been made into nanoparticles. Meanwhile, students will be trained in working with these exotic materials.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Expected Waste Streams Originating from the Decommissioning of NPPs: German Situation as Compared to International Standards - 16447

This paper presents an overview of current and future decommissioning projects of NPPs in Germany. The expected amounts of residual radioactive materials are given. Their possible utilization and disposal routes, which are stipulated by the German Radiation Protection Ordinance, are presented and compared to those resulting from the French zoning model. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Decommissioning and Demolition of a Redundant UK Research Facility at AWE Aldermaston - 12453

The redundant two-storey brick built research facility on the AWE Site at Aldermaston, UK is in the closing stages of decommissioning and demolition. The facility was used for a variety of purposes up to 1995 predominately involving the use of alpha-emitting isotopes. The two main areas of alpha-based contamination have been decommissioned with the removal of hot -boxes and fume cupboards on the ground floor and HEPA filter units and ventilation equipment on the first floor. Many of these activities were undertaken using both airline fed suits, (supplied via a free standing mobile unit), and full face respirators. Asbestos materials were located and cleared from the first floor by specialist contractor. All sections of active drain running from the building to the site active effluent disposal system were removed early in the program using established techniques with specialist monitoring equipment used to provide confidence in the data required for disposal of the decommissioning debris. In particular a dedicated High Resolution Gamma Spectrometer (radioactive materials scanning unit) was utilized to categorise waste drums and wrapped packages. The building has been decommissioned and the monitoring and sampling of the structure was completed in November 2011 - the results demonstrating that the building was clear of contamination in accordance with UK clearance and exemption requirements. The demolition plan was developed and implemented in December with site excavation of foundations and site clearance currently ongoing in preparation for final site backfill activities and project close. A number of useful lessons have been learnt during the operations and are set out at the rear of the main text. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Optimization of the Shielding of a Canister Loading Machine with the deterministic Code Attila{sup R} - Paper 61

In general shielding components are designed with very conservative assumptions and boundary conditions to cover all possibly occurring situations. This often results in an overestimated shielding design in a lot of different fields of application; the shielding of components will be heavier and more expensive than needed. In most cases an initial shielding calculation in the design process shows an unbalanced shielding effect of the shielded radiation source. Sometimes the dose rate limits are slightly exceeded in several areas. From this arises a need for an optimization of the shielding structure. The goal is to decrease the radiation level in the accessible areas to keep the dose rate limits and to reduce the amount of the used shielding material if possible. Shielding material is moved from areas with higher shielding factors to which with lower ones. The optimization process is shown in an example of a loading machine for vitrified waste in canisters. The design review with an optimization of the shielding geometry was done with the deterministic code Attila{sup R} of the company Transpire, Inc. Attila{sup R} is a powerful tool to calculate and optimize the shielding effect of handling facilities for radioactive materials. It is easy to handle with its CAD integration module and an easy setup of the calculation cases. Areas with a low shielding effect which results in an exceeding of the dose rate limits can be found quickly with the illustration due to an Iso-Dose-Rate-Surface in the 3D Model. In these areas the shielding effect has to be improved. Otherwise there is material, which has no significant influence on the shielding effect and makes the handling facilities heavier and more expensive as needed. The optimization of the shielding design of the canister loading machine in the current case results in lower dose rates outside the facility to keep the dose rate limits. The demand of shielding material could be reduced and resulted in a cost reduction on the facility itself and furthermore in the caring facilities to transport this facility (cranes, etc.). (authors)

42 ENGINEERING↗
DOE OSTI2012

Tracking and Monitoring with Dosimeter-Enabled ARG-US RFID System - 12009

Automated monitoring and tracking of materials with radio frequency identification (RFID) technology can significantly improve both the operating efficiency of radiological facilities and the application of the ALARA (as low as reasonably achievable) principle in them. One such system, called ARG-US, has been developed by Argonne National Laboratory for the U.S. Department of Energy (DOE) Packaging and Certification Program to use in managing sensitive nuclear and radioactive materials. Several ARG-US systems are in various stages of deployment and advanced testing across DOE sites. ARG-US utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users. In conjunction with global positioning system (GPS) tracking provided by TRANSCOM, the system can also monitor and track packages during transport. A compact dosimeter has been incorporated in the ARG-US tags via an onboard universal asynchronous receiver/transmitter interface. The detector has a wide measurement range for gamma radiation - from 0.1 mSv/h to 8 Sv/h. The detector is able to generate alarms for both high and low radiation and for a high cumulative dose. In a large installation, strategically located dosimeter-enabled tags can yield an accurate, real-time, 2D or 3D dose field map that can be used to enhance facility safety, security, and safeguards. This implementation can also lead to a reduced need for manned surveillance and reduced exposure of personnel to radiation, consistent with the ALARA principle at workplaces. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗