Uranium Aerosol Research Articles

Uranium Body Clearance Kinetics-A Long-term Follow-up Study of Retired Nuclear Fuel Workers.

Nuclear industry workers exposed to uranium aerosols may risk kidney damage and radiation-induced cancer. This warrants the need for well-established dose and risk assessments, which can be greatly improved by using material-specific absorption parameters in the ICRP Human Respiratory Tract Model. The present study focuses on the evaluation of the slow dissolution rate ( s s , d -1 ), a parameter that is difficult to quantify with in vitro dissolution studies, especially for more insoluble uranium compounds. A long-term follow-up of urinary excretion after the cessation of chronic inhalation exposure can provide a better estimate of the slow-rate dissolution. In this study, two workers, previously working for >20 y at a nuclear fuel fabrication plant, provided urine samples regularly for up to 6 y. One individual had worked at the pelletizing workshop with the known presence of uranium dioxide (UO 2 ) and triuranium octoxide (U 3 O 8 ). The second individual worked at the conversion workshop where multiple compounds, including uranium hexafluoride (UF 6 ), uranium dioxide (UO 2 ), ammonium uranyl carbonate, and AUC [UO 2 CO 3 ·2(NH 4 ) 2 CO 3 ], are present. Data on uranium concentration in urine during working years were also available for both workers. The daily excretion of uranium by urine was characterized by applying non-linear least square regression fitting to the urinary data. Material-specific parameters, such as the activity median aerodynamic diameter (AMAD), the respiratory tract absorption parameters, rapid fraction ( f r ,), rapid dissolution rate ( s r , d -1 ), and slow dissolution rate ( s s , d -1 ) and alimentary tract transfer factor ( f A ) acquired from previous work along with default absorption types, were applied to urine data, and the goodness of fit was evaluated. Thereafter intake estimates and dose calculations were performed. For the ex-pelletizing worker, a one-compartment model with a clearance half-time of 662 ± 100 d ( s s = 0.0010 d -1 ) best represented the urinary data. For the ex-conversion worker, a two-compartment model with a major [93% of the initial urinary excretion (A 0 )] fast compartment with a clearance half-time of 1.3 ± 0.4 d ( s r = 0.5 d -1 ) and a minor (7% of A 0 ) slow compartment with a half-time of 394 ± 241 d ( s s = 0.002 d -1 ) provided the best fit. The results from the data-fitting of urinary data to biokinetic models for the ex-conversion worker demonstrated that in vitro derived experimental parameters (AMAD = 20 μm, f r = 0.32, s r = 27 d -1 , s s = 0.0008 d -1 , f A = 0.005) from our previous work best represented the urinary data. This resulted in an estimated intake rate of 0.66 Bq d -1 . The results from the data-fitting of urinary data to biokinetic models for the ex-pelletizing worker indicated that the experimental parameters (AMAD = 10 μm and 20 μm, f r = 0.008, s r = 12 d -1 , f A = 0.00019) from our previous dissolution studies with the slow rate parameter step-wise optimized to urine-data ( s s = 0.0008 d -1 ) gave the best fit. This resulted in an estimated intake rate of 5 Bq d -1 . Experimental parameters derived from in vitro dissolution studies provided the best fit for the subject retired from work at the conversion workshop, where inhalation exposure to a mix of soluble (e.g., AUC, UF 6 ) and relatively insoluble aerosol (e.g., UO 2 ) can be assumed. For the subject retired from work at the pelletizing workshop, which involved exposure to relatively insoluble aerosols (UO 2 and U 3 O 8 ), a considerably higher s s than obtained in dissolution studies provided a better representation of the urinary data and was comparable to reported s s values for UO 2 and U 3 O 8 in other studies. This implies that in vitro dissolution studies of insoluble material can be uncertain. When evaluating the results from the retrospective fitting of urine data, it is evident that the urine samples acquired after cessation of exposure provide less fluctuation. Long-term follow-up of uranium excretion after cessation of exposure is a good alternative for determining absorption parameters and can be considered the most viable way for determining the slow rate for more insoluble material.

СРАВНЕНИЕ РАДИАЦИОННОЙ И ХИМИЧЕСКОЙ ТОКСИЧНОСТИ СОЕДИНЕНИЙ УРАНА НА ОСНОВЕ РАСЧЕТА ПО НОВЫМ БИОКИНЕТИЧЕСКИМ МОДЕЛЯМ МКРЗ

Purpose: Harmonization and improvement of the system for regulating the internal radiation exposure of workers and the basic requirements for ensuring radiation safety, taking into account the application of new international requirements and recommendations. Material and methods: The article presents a comparison of the radiation and chemical toxicity of uranium compounds, obtained on the basis of calculating the levels of inhalation intake and committed effective dose depending on the types of compounds F, M, F/M and M/S in the AMAD range from 0.3 to 20 μm for typical isotopic compositions of natural (NU), depleted (DU), low enriched (LEU) and highly enriched (HEU) uranium, which lead to the maximum permissible concentration of uranium in the kidneys. The calculations were carried out using new ICRP biokinetic models, which give more physiologically realistic representations of uptake and retention in organs and tissues, and excretion. Results: The dynamics of uranium activity in the kidneys was calculated for constant chronic inhalation intake over a 50-year period and for acute intake. It was shown that in case of chronic intake, the rate of accumulation of uranium in the kidneys, expressed in relative units, does not depend on the AMAD in the range from 0.3 to 20 μm and slightly depends on the types of compounds F, F/M, M and M/S, which include almost all chemical compounds of uranium. In case of acute intake, there is a rapid, within 1–3 days, an increase of uranium in the kidneys to a maximum value and then a gradual decrease to a value of 20 % of the maximum value in 20–60 days, depending on the type of compound F, M, F/M, M/S and AMAD in a wide range of values from 0.3 to 20 µm. To compare the radiation and chemical toxicity of uranium, the values of the committed effective dose were calculated, which is formed after intake of uranium aerosols of the types F, M, F/M and M/S and AMADs from 0.3 to 20 µm in an amount that creates the maximum concentration of uranium in the kidneys 0.3 µg/g for chronic intake and 3 µg/g for acute intake. The values of uranium intake per year in milligrams, which form the maximum concentration of uranium in the kidneys of 0.3 µg/g, in case of constant chronic intake of uranium aerosols, as well as the values of uranium intake in milligrams, which form the maximum concentration of uranium in the kidneys of 3 μg/g after a single intake of uranium aerosols in both case of the types F, M, F/M and M/S and AMAD in the range from 0.3 to 20 µm were calculated, which are evidently independent of the considered isotopic composition of the uranium. Conclusion: It is shown that chemical toxicity prevails over radiation toxicity for the types of uranium compounds F and F/M for all considered uranium isotopic composition, except for HEU; for the type of compound M it is the same for mixtures of NU and DU, and for the type M/S radiation toxicity prevails for all considered uranium isotopic composition. In case of chronic intake at committed effective dose exposure rate of several mSv per year, workers can suffer from the chemical toxicity of uranium when working with F and F/M compounds and isotope compositions of natural (NU), depleted (DU) and low enriched (LEU) uranium already after 1–2 years of work. In case of acute intake, the chemical toxicity of uranium should be taken as a criterion for limiting exposure for compounds F and F/M, and also partially M (for uranium isotope compositions of NU, DU and LEU), which can significantly, by tens and hundreds of times, reduce the permissible limit of uranium intake.

Wounding characteristics and treatment principles of ground anti-armored vehicle ammunition against armored crew

The wound mechanism, injury characteristics and treatment principles of anti-armored vehicle ammunition against armored crew in the past 20 years are summarized in this paper. Shock vibration, metal jet, depleted uranium aerosol and post armor breaking effect are the main factors for wounding armored crew. Their prominent characteristics are severe injury, high incidence of bone fracture, high rate of depleted uranium injury, and high incidence of multiple/combined injuries. During the treatment, attention must be paid on that the space of armored vehicle is limited, and the casualties should be moved outside of the cabin for comprehensive treatment. Especially, the management of depleted uranium injury and burn/inhalation injury are more important than other injuries for the armored wounds.

Study on the Equivalence of Metallic-Cerium-Simulated Uranium-Aerosol Generation under Fire

Uranium aerosols are released from uranium-containing materials in high-temperature environments caused by nuclear accidents or other processes. Research on the generation characteristics of uranium aerosols under such conditions is an important part of nuclear-safety analysis. In this experiment, the similarity between metal cerium aerosols and uranium material aerosols was evaluated from the aspects of particle size distribution and source term. Combined with the experiment data, the effect of air flow rate and sampling time is discussed. The calculation result of the air release fraction (ARF) is 6.07 × 10−3–4.8 × 10−2, and the respirable fraction (RF) is 0.810–0.978, respectively, showing that the size distribution of particles and ARF of the cerium aerosol are different from the results of the uranium aerosols in the literature, while the RF is similar to the results obtained by using the uranium–niobium alloy in the literature.

Particle Size Dependent Dissolution of Uranium Aerosols in Simulated Gastrointestinal Fluids.

Uranium aerosol exposure can be a health risk factor for workers in the nuclear fuel industry. Good knowledge about aerosol dissolution and absorption characteristics in the gastrointestinal tract is imperative for solid dose assessments and risk management. In this study, an in vitro dissolution model of the GI tract was used to experimentally study solubility of size-fractionated aerosols. The aerosols were collected from four major workshops in a nuclear fuel fabrication plant where uranium compounds such as uranium hexafluoride (UF 6 ), uranium dioxide (UO 2 ), ammonium uranyl carbonate, AUC [UO 2 CO 3 ·2(NH 4 ) 2 CO 3 ] and triuranium octoxide (U 3 O 8 ) are present. The alimentary tract transfer factor, f A , was estimated for the aerosols sampled in the study. The transfer factor was derived from the dissolution in the small intestine in combination with data on absorption of soluble uranium. Results from the conversion workshop indicated a f A in line with what is recommended (0.004) by the ICRP for inhalation exposure to Type M materials. Obtained transfer factors, f A , for the powder preparation and pelletizing workshops where UO 2 and U 3 O 8 are handled are lower for inhalation and much lower for ingestion than those recommended by the ICRP for Type M/S materials f A = 0.00029 and 0.00016 vs. 0.0006 and 0.002, respectively. The results for ingestion and inhalation f A indicate that ICRP's conservative recommendation of f A for inhalation exposure is applicable to both ingestion and inhalation of insoluble material in this study. The dissolution- and subsequent absorption-dependence on particle size showed correlation only for one of the workshops (pelletizing). The absence of correlation at the other workshops may be an effect of multiple chemical compounds with different size distribution and/or the reported presence of agglomerated particles at higher cut points having more impact on the dissolution than particle size. The impact on dose coefficients [committed effective dose (CED) per Bq] of using experimental f A vs. using default f A recommended by the ICRP for the uranium compounds of interest for inhalation exposure was not significant for any of the workshops. However, a significant impact on CED for ingestion exposure was observed for all workshops when comparing with CED estimated for insoluble material using ICRP default f A . This indicates that the use of experimentally derived site-specific f A can improve dose assessments. It is essential to acquire site-specific estimates of the dissolution and absorption of uranium aerosols as this provides more realistic and accurate dose- and risk-estimates of worker exposure. In this study, the results indicate that ICRP's recommendations for ingestion of insoluble material might overestimate absorption and that the lower f A found for inhalation could be more realistic for both inhalation and ingestion of insoluble material.

ПРИМЕНЕНИЕ НОВЫХ РЕКОМЕНДАЦИЙ МКРЗ ДЛЯ РАСЧЕТА ДОЗ ПЕРСОНАЛА ПРИ ИНГАЛЯЦИОННОМ ПОСТУПЛЕНИИ РАДИОНУКЛИДОВ УРАНА

Purpose: Harmonization and improvement of the system for regulating the internal radiation exposure of workers and the basic requirements for ensuring radiation safety, taking into account the application of new international requirements and recommendations. Material and methods: A brief description of the procedure for calculating absorbed and equivalent doses in organs and tissues after the intake of radionuclides into the human body is presented, using the biokinetic and dosimetric models adopted in the new ICRP recommendations, as well as a discussion of the impact of these changes on the results of calculating dose coefficients for the case of inhalation intake of uranium-235 radionuclide. Results: The effective dose values and equivalent doses to organs and tissues for workers were calculated depending on the time after a single inhalation intake of uranium-235 aerosol, according to new models [1–12] and according to previous ICRP models [15, 16]. The calculation of the effective dose according to the new models included calculations of equivalent doses for 14 main organs and tissues and 13 organs and tissues classified as “remainder tissues” as described in ICRP Publication 103 [3]. The committed effective dose was then calculated according to the new approach using the average of the equivalent doses for the reference adult male, HTM, and the reference adult female, HTF, as well as the tissue and organ weighting factors, WT, adopted in ICRP Publication 103. The values of the effective dose and equivalent doses on the red bone marrow, lungs and remainder tissues vs time in the range from several days to 18250 days (50 years) after a single inhalation intake of an aerosol of uranium-235 for standard value AMAD=5 µm and types of compounds F, M, S, F/M and M/S are presented according to new and previous ICRP models. It is shown that the value of the dose coefficient for type F, calculated by new models, is 2.6 times (2.3E-07÷6.0E-07) less than that calculated by previous ICRP models, and the value of the dose coefficient for type F/M calculated by new models is 1.6 times (3.8E-07÷6.0E-07) less than the value of the dose coefficient for type F calculated by previous ICRP models. For uranium trioxide UO3, taking into account its transition from compound type M to F/M, the value of the dose coefficient for committed effective dose according to the updated model of the respiratory tract is 4.7 times (3.8E-07÷1.8E-06) less than the corresponding value for the previous model of the type M respiratory tract. The committed effective dose value for compound type M, calculated using the new models, is 1.4 times (1.3E-06÷1.8E-06) less than the same value calculated using the previous ICRP models. The value of the committed effective dose for type M/S compounds (which, according to the new model of the respiratory tract, include uranium oxide U3O8 and dioxide UO2), calculated according to new models, is 1.2 times (5.1E-06÷6.1E-06) less than the value calculated from previous ICRP models for type S compounds (which included U3O8 and UO2 in the previous respiratory tract model). Conclusion: From the above data it follows that in case of the adoption of national radiation safety standards to new ICRP models, differences in the values of dose coefficients will result in a change of annual limits of intake (ALI) in the corresponding proportion for the types of uranium aerosol compounds noted above.

Particle Size-dependent Dissolution of Uranium Aerosols in Simulated Lung Fluid: A Case Study in a Nuclear Fuel Fabrication Plant.

Inhalation exposure to uranium aerosols can be a concern in nuclear fuel fabrication. The ICRP provides default absorption parameters for various uranium compounds but also recommends determination of material-specific absorption parameters to improve dose calculations for individuals exposed to airborne radioactivity. Aerosol particle size influences internal dosimetry calculations in two potentially significant ways: the efficiency of particle deposition in the various regions of the respiratory tract is dependent on aerodynamic particle size, and the dissolution rate of deposited materials can vary according to particle size, shape, and porosity because smaller particles tend to have higher surface-to-volume ratios than larger particles. However, the ICRP model assumes that deposited particles of a given material dissolve at the same rate regardless of size and that uptake to blood of dissolved material normally occurs instantaneously in all parts of the lung (except the anterior portion of the nasal region, where zero absorption is assumed). In the present work, the effect of particle size on dissolution in simulated lung fluid was studied for uranium aerosols collected at the plant, and its influence on internal dosimetry calculations was evaluated. Size fractionated uranium aerosols were sampled at a nuclear fuel fabrication plant using portable cascade impactors. Absorption parameters, describing dissolution of material according to the ICRP Human Respiratory Tract Model, were determined in vitro for different size fractions using simulated lung fluid. Samples were collected at 16 time-points over a 100-d period. Uranium content of samples was determined using inductively coupled plasma mass spectrometry and alpha spectrometry. In addition, supplementary experiments to study the effect of pH drift and uranium adsorption on filter holders were conducted as they could potentially influence the derived absorption parameters. The undissolved fraction over time was observed to vary with impaction stage cut-point at the four main workshops at the plant. A larger fraction of the particle activity tended to dissolve for small cut-points, but exceptions were noted. Absorption parameters (rapid fraction, rapid rate, and slow rate), derived from the undissolved fraction over time, were generally in fair agreement with the ICRP default recommendations for uranium compounds. Differences in absorption parameters were noted across the four main workshops at the plant (i.e., where the aerosol characteristics are expected to vary). The pelletizing workshop was associated with the most insoluble material and the conversion workshop with the most soluble material. The correlation between derived lung absorption parameters and aerodynamic particle size (impactor stage cut-point) was weak. For example, the mean absorption parameters derived from impaction stages with low (taken to be <5 μm) and large (≥5 μm) cut-points did not differ significantly. Drift of pH and adsorption on filter holders appeared to be of secondary importance, but it was found that particle leakage can occur. Undissolved fractions and to some degree derived lung absorption parameters were observed to vary depending on the aerodynamic size fraction studied, suggesting that size fractionation (e.g., using cascade impactors) is appropriate prior to conducting in vitro dissolution rate experiments. The 0.01-0.02 μm and 1-2 μm size ranges are of particular interest as they correspond to alveolar deposition maxima in the Human Respiratory Tract Model (HRTM). In the present work, however, the dependency on aerodynamic size appeared to be of minor importance, but it cannot be ruled out that particle bounce obscured the results for late impaction stages. In addition, it was noted that the time over which simulated lung fluid samples are collected (100 d in our case) influences the curve-fitting procedure used to determine the lung absorption parameters, in particular the slow rate that increased if fewer samples were considered.

Simulation of particle size distribution and source term for uranium aerosols based on zinc generated under fire conditions

Fire is a typical scenario for nuclear accidents, in which uranium-containing materials will release uranium aerosols. Limited by many aspects, uranium aerosols cannot be used directly to carry out nuclear emergency training and equipment development research under real conditions. Therefore, it is necessary to filter and evaluate available surrogate materials through experiments. In this paper, metallic zinc is used to study the source term and particle size distribution of zinc aerosols under fire conditions, and the feasibility of zinc as a surrogate material for uranium aerosol research is evaluated based on the data of uranium aerosols in the literature. The experimental results show that the mass median aerodynamic diameter obtained by fitting with lognormal distribution is between 0.36 and 0.59 µm. The cumulative mass fraction distribution data are consistent with those given in the literature, indicating that the zinc aerosols generated under fire have a nice simulation effect on uranium aerosols in terms of particle size distribution. As for air release fraction (ARF) and respirable fraction (RF), the experiment of zinc aerosols is greater than that of uranium aerosols, which means the simulation in ARF and RF is not so good.

Uranium Aerosol Activity Size Distributions at a Nuclear Fuel Fabrication Plant.

Inhalation of uranium aerosols is a concern in nuclear fuel fabrication. Determination of committed effective doses and lung equivalent doses following inhalation intake requires knowledge about aerosol characteristics; e.g., the activity median aerodynamic diameter (AMAD). Cascade impactor sampling of uranium aerosols in the breathing zone of nuclear operators was carried out at a nuclear fuel fabrication plant producing uranium dioxide via ammonium uranyl carbonate. Complementary static sampling was carried out at key process steps. Uranium on impaction substrates was measured using gross alpha counting and alpha spectrometry. Activity size distributions were evaluated for both unimodal and bimodal distributions. When a unimodal distribution was assumed, the average AMAD in the operator breathing zone at the workshops was 12.9–19.3 μm, which is larger than found in previous studies. Certain sampling occasions showed variable isotope ratios (234U/238U) at different impactor stages, indicating more than one population of particles; i.e., a multimodal activity size distribution. When a bimodal distribution (coarse and fine fraction) was assumed, 75–88% of the activity was associated with an AMAD of 15.2–18.9 μm (coarse fraction). Quantification of the AMAD of the fine fraction was associated with large uncertainties. Values of 1.7–7.1 μm were obtained. Static sampling at key process steps in the workshops showed AMADs of 4.9–17.2 μm, generally lower than obtained by breathing zone sampling, when a unimodal distribution was assumed. When a bimodal distribution was assumed, a smaller fraction of the activity was associated with the coarse fraction compared to breathing zone sampling. This might be due to impactor positioning during sampling and sedimentation of large particles. The average committed effective dose coefficient for breathing zone sampling and a bimodal distribution was 1.6–2.6 μSv Bq−1 for 234U when Type M/S absorption parameters were assumed (5.0 μSv Bq−1 for an AMAD of 5 μm). The corresponding lung equivalent dose coefficient was 3.6–10.7 μSv Bq−1 (29.9 μSv Bq−1 for an AMAD of 5 μm). The predicted urinary excretion level 100 d after inhalation intake was found to be 13-34% of that corresponding to an AMAD of 5 μm. Uranium aerosols generated at a nuclear fuel fabrication plant using ammonium uranyl carbonate route of conversion were associated with larger AMADs compared to previous work, especially when sampling of aerosols was carried out in the operator breathing zone. A bimodal activity size distribution can be used in calculations of committed effective doses and lung equivalent doses, but parameters associated with the fine fraction must be interpreted with care due to large uncertainties.

Latest improvements in isotopic uranium particle analysis by large geometry–secondary ion mass spectrometry for nuclear safeguards purposes

Large geometry secondary ion mass spectrometry can be efficiently used to analyze uranium aerosol particles from dust samples in the search for undeclared nuclear activities. Automated sample screening measurements are followed by more precise and accurate microbeam measurements of both the major and minor uranium isotopes on selected individual particles. The quality of this work is essential in order to be able to draw valuable safeguards conclusions. This paper describes the latest developments that have been undertaken to enhance the detection limits and to reduce the uranium isotope measurement uncertainty. It includes improvements in the analytical protocol as well as in the instrument acquisition software and data reduction method. Recent useful yield measurements have been performed on uranium monodispersed particles using different primary bombardment conditions to compare to previously obtained data. Comparison of uranium isotope measurements when using pyrolytic graphite or silica planchets as a sample substrate will also be presented.

AEROSOL SIZE DISTRIBUTION IN URANIUM PROCESSING FACILITY FOR APPLICATION IN THE ASSESSMENT OF INTERNAL EXPOSURE.

In circumstances of inhalation exposure the radiotoxicity of radionuclide depends on the fractional deposition of inhaled activity, which is governed by the aerodynamic characteristics of the particles. Although International Commission on Radiological Protection (ICRP) provides default size distribution parameters for work place aerosols, inhaled aerosol characteristics may be different depending on the physical and chemical conditions of aerosol generation process. The present study is undertaken to determine the particle activity-size distribution during operation of various significant processes of uranium metal production facility of Bhabha Atomic Research Centre (BARC). Activity median aerodynamic diameter (AMAD) and its geometric standard deviation (GSD) were estimated for these process areas for uranium aerosol particles. The AMAD varied from 3.2 to 10.06 µm and GSD ranged from 1.5 to 3.0. Also, the obtained size distribution was tested by Pearson's chi-squared distribution test method to ascertain the assumption that the particle size distribution is best described by log-normal relationship.

Uranium aerosols at a nuclear fuel fabrication plant: Characterization using scanning electron microscopy and energy dispersive X-ray spectroscopy

Detailed aerosol knowledge is essential in numerous applications, including risk assessment in nuclear industry. Cascade impactor sampling of uranium aerosols in the breathing zone of nuclear operators was carried out at a nuclear fuel fabrication plant. Collected aerosols were evaluated using scanning electron microscopy and energy dispersive X-ray spectroscopy. Imaging revealed remarkable variations in aerosol morphology at the different workshops, and a presence of very large particles (up to≅100×50μm2) in the operator breathing zone. Characteristic X-ray analysis showed varying uranium weight percentages of aerosols and, frequently, traces of nitrogen, fluorine and iron. The analysis method, in combination with cascade impactor sampling, can be a powerful tool for characterization of aerosols. The uranium aerosol source term for risk assessment in nuclear fuel fabrication appears to be highly complex.

Adaptive Visual Sort and Summary of Micrographic Images of Nanoparticles for Forensic Analysis.

Image classification of nanoparticles from scanning electron microscopes for nuclear forensic analysis is a long, time consuming process. Months of analyst time may initially be required to sift through images in order to categorize morphological characteristics associated with nanoparticle identification. Subsequent assessment of newly acquired images against identified characteristics can be equally time consuming. We present INStINCt, our Intelligent Signature Canvas, as a framework for quickly organizing image data in a web-based canvas framework that partitions images based on features derived from convolutional neural networks. This work is demonstrated using particle images from an aerosol study conducted by Pacific Northwest National Laboratory under the auspices of the U.S. Army Public Health Command to determine depleted uranium aerosol doses and risks.

Characterization of Respirable Uranium Aerosols from Various Uranium Alloys in Fire Events

Aerosols dispersed from the oxidation of various uranium alloys exposed to air and direct flame impingement from combustible substrates are characterized. An apparatus was designed to incorporate desired characteristics of previous experiments on uranium to sample aerosol on a kilogram scale in a laboratory environment. Previous studies involving β-phase stabilized uranium (99.25 wt% U:0.75 wt% Ti) were benchmarked using an identical alloy with identical characteristics of the original specimens. Other studies involving a-phase uranium (100 wt% U) were also benchmarked in this experiment. Unique to this study is the use of γ-phase stabilized uranium (94 wt% U:6 wt% Nb). These three alloys represent the crystallographic range of typical uranium metals, providing a complete spectrum of potential uranium aerosolization. Oxidation rates and extents observed in these experiments were directly comparable to existing data and provided correlation between previous studies. These experiments indicate a distinct or...

Aerosol size distribution in a uranium processing and fuel fabrication facility

In the nuclear fuel complex, magnesium diuranate is processed to produce UO(2) through different chemical and metallurgical processes. UO(2) powder is compacted to produce uranium pallets as fuel. International Commission on Radiological Protection has considered default particle size of 5-mum activity median aerodynamic diameter (AMAD) and 2.5 of geometric standard deviation (GSD) for working out dose coefficients. There is a likelihood of variation in the particle size during each stage of operation. The present study is undertaken to determine the prevailing uranium aerosol size distribution at every stage of operation using Anderson impactor with glass fibre filter paper as collection substrate. AMAD and respective GSD were determined. Aerosol size distribution was studied. Airborne uranium concentration was found to be higher for higher particle sizes in all areas. Average AMAD for different locations varied from 5.8 to 7.7 mum with GSD from 1.63 to 6.73 and the ratio of calculated ALI to standard varies from 1.13 to 1.55.

PHYSICOCHEMICAL CHARACTERIZATION OF CAPSTONE DEPLETED URANIUM AEROSOLS II: PARTICLE SIZE DISTRIBUTIONS AS A FUNCTION OF TIME

The Capstone Depleted Uranium (DU) Aerosol Study, which generated and characterized aerosols containing DU from perforation of armored vehicles with large-caliber DU penetrators, incorporated a sampling protocol to evaluate particle size distributions. Aerosol particle size distribution is an important parameter that influences aerosol transport and deposition processes as well as the dosimetry of the inhaled particles. These aerosols were collected on cascade impactor substrates using a pre-established time sequence following the firing event to analyze the uranium concentration and particle size of the aerosols as a function of time. The impactor substrates were analyzed using proportional counting, and the derived uranium content of each served as input to the evaluation of particle size distributions. Activity median aerodynamic diameters (AMADs) of the particle size distributions were evaluated using unimodal and bimodal models. The particle size data from the impactor measurements were quite variable. Most size distributions measured in the test based on activity had bimodal size distributions with a small particle size mode in the range of between 0.2 and 1.2 microm and a large size mode between 2 and 15 microm. In general, the evolution of particle size over time showed an overall decrease of average particle size from AMADs of 5 to 10 microm shortly after perforation to around 1 microm at the end of the 2-h sampling period. The AMADs generally decreased over time because of settling. Additionally, the median diameter of the larger size mode decreased with time. These results were used to estimate the dosimetry of inhaled DU particles.

CAPSTONE DEPLETED URANIUM AEROSOL BIOKINETICS, CONCENTRATIONS, AND DOSES

One of the principal goals of the Capstone Depleted Uranium (DU) Aerosol Study was to quantify and characterize DU aerosols generated inside armored vehicles by perforation with a DU penetrator. This study consequently produced a database in which the DU aerosol source terms were specified both physically and chemically for a variety of penetrator-impact geometries and conditions. These source terms were used to calculate radiation doses and uranium concentrations for various scenarios as part of the Capstone Human Health Risk Assessment (HHRA). This paper describes the scenario-related biokinetics of uranium, and summarizes intakes, chemical concentrations to the organs, and E(50) and HT(50) for organs and tissues based on exposure scenarios for personnel in vehicles at the time of perforation as well as for first responders. For a given exposure scenario (duration time and breathing rates), the range of DU intakes among the target vehicles and shots was not large, about a factor of 10, with the lowest being for a ventilated operational Abrams tank and the highest being for an unventilated Abrams with DU penetrator perforating DU armor. The ranges of committed effective doses were more scenario-dependent than were intakes. For example, the largest range, a factor of 20, was shown for scenario A, a 1 min exposure, whereas, the range was only a factor of two for the first-responder scenario (E). In general, the committed effective doses were found to be in the tens of mSv. The risks ascribed to these doses are discussed separately.

PHYSICOCHEMICAL CHARACTERIZATION OF CAPSTONE DEPLETED URANIUM AEROSOLS IV: IN VITRO SOLUBILITY ANALYSIS

As part of the Capstone Depleted Uranium (DU) Aerosol Study, the solubility of selected aerosol samples was measured using an accepted in vitro dissolution test system. This static system was employed along with a SUF (synthetic ultrafiltrate) solvent, which is designed to mimic the physiological chemistry of extracellular fluid. Using sequentially obtained solvent samples, the dissolution behavior over a 46-d test period was evaluated by fitting the measurement data to two- or three-component negative exponential functions. These functions were then compared with Type M and S absorption taken from the International Commission on Radiological Protection Publication 66 Human Respiratory Tract Model. The results indicated that there was a substantial variability in solubility of the aerosols, which in part depended on the type of armor being impacted by the DU penetrator and the particle size fraction being tested. Although some trends were suggested, the variability noted leads to uncertainties in predicting the solubility of other DU-based aerosols. Nevertheless, these data provide a useful experimental basis for modeling the intake-dose relationships for inhaled DU aerosols arising from penetrator impact on armored vehicles.

CALCULATING CAPSTONE DEPLETED URANIUM AEROSOL CONCENTRATIONS FROM BETA ACTIVITY MEASUREMENTS

Beta activity measurements were used as surrogate measurements of uranium mass in aerosol samples collected during the field testing phase of the Capstone Depleted Uranium (DU) Aerosol Study. These aerosol samples generated by the perforation of armored combat vehicles were used to characterize the DU source term for the subsequent Human Health Risk Assessment (HHRA) of Capstone aerosols. Establishing a calibration curve between beta activity measurements and uranium mass measurements is straightforward if the uranium isotopes are in equilibrium with their immediate short-lived, beta-emitting progeny. For DU samples collected during the Capstone study, it was determined that the equilibrium between the uranium isotopes and their immediate short-lived, beta-emitting progeny had been disrupted when penetrators had perforated target vehicles. Adjustments were made to account for the disrupted equilibrium and for wall losses in the aerosol samplers. Values for the equilibrium fraction ranged from 0.16 to 1, and the wall loss correction factors ranged from 1 to 1.92. This paper describes the process used and adjustments necessary to calculate uranium mass from proportional counting measurements.

OVERVIEW OF THE CAPSTONE DEPLETED URANIUM STUDY OF AEROSOLS FROM IMPACT WITH ARMORED VEHICLES: TEST SETUP AND AEROSOL GENERATION, CHARACTERIZATION, AND APPLICATION IN ASSESSING DOSE AND RISK

The Capstone Depleted Uranium (DU) Aerosol Characterization and Risk Assessment Study was conducted to generate data about DU aerosols generated during the perforation of armored combat vehicles with large-caliber DU penetrators, and to apply the data in assessments of human health risks to personnel exposed to these aerosols, primarily through inhalation, during the 1991 Gulf War or in future military operations. The Capstone study consisted of two components: 1) generating, sampling, and characterizing DU aerosols by firing at and perforating combat vehicles, and 2) applying the source-term quantities and characteristics of the aerosols to the evaluation of doses and risks. This paper reviews the background of the study including the bases for the study, previous reviews of DU particles and health assessments from DU used by the U.S. military, the objectives of the study components, the participants and oversight teams, and the types of exposures it was intended to evaluate. It then discusses exposure scenarios used in the dose and risk assessment and provides an overview of how the field tests and dose and risk assessments were conducted.