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As part of an epidemiological study, doses from intake of radionuclides were estimated for workers employed during a 52-year period at the Rocketdyne/Atomics International facility in California. The facility was involved in a variety of research programmes, including nuclear fuel fabrication, spent nuclear fuel decladding, and reactor operation and disassembly. Most of the documented intakes involved inhalation of enriched uranium (U), fission products, or plutonium (Pu). Highest doses were estimated for a group of workers exposed to airborne uranium aluminide (UAl(x)) during the fabrication of reactor fuel plates. Much of the exposure to UAl(x) occurred early in the fuel fabrication programme, before it was recognised that intake and lung retention were being underestimated from urinary data due to an unexpected delayed dissolution of the inhaled material. In workers who had been removed from exposure, the rate of urinary excretion of U increased for a few months, peaked, and then declined at a rate consistent with moderately soluble material. This pattern differs markedly from the monotonically decreasing absorption rates represented by the default absorption types in the Human Respiratory Tract Model (HRTM) of the International Commission on Radiological Protection (ICRP). This paper summarises the findings on the behaviour of UAl(x) in these workers and describes material-specific parameter values of the HRTM based on this information.
Values for the absorption parameters were compared after inhalation or intratracheal instillation of 1.5 microm mass median aerodynamic diameter (MMAD) 233U3O8 particles into the lungs of HMT strain rats. The two sets of parameter values were similar, as were the calculated dose coefficients and predicted biokinetics for workers. Hence the inhalation and instillation techniques can probably both be used to generate values of the absorption parameters for U3O8.
This paper provides guidance on the most appropriate monitoring procedures and intervals, the likely uncertainties in the assessment of intake and recommendations on appropriate investigation levels for repeated exposures to uranium trioxide, octoxide and dioxide of natural composition.
Methods have been developed to assess the size distribution of alpha emitting particles of reactor fuel of known composition captured on air sampler filters. The sizes of uranium oxide and plutonium oxide particles were determined using a system based on CR-39 solid-state nuclear track detectors. The CR-39 plastic was exposed to the deposited particles across a 400 microm airgap. The exposed CR-39 was chemically etched to reveal clusters of tracks radially dispersed from central points. The number and location of the tracks were determined using an optical microscope with an XY motorised table and image analysis software. The sample mounting arrangement allowed individual particles to be simultaneously viewed with their respective track cluster. The predicted diameters correlated with the actual particle diameters, as measured using the optical microscope. The efficacy of the technique was demonstrated with particles of natural uranium oxide (natUO2) of known size, ranging from 4 to 150 microm in diameter. Two personal air sampler (PAS) filters contaminated with actinide particles were placed against CR-39 and estimated to have size distributions of 0.8 and 1.0 microm activity median aerodynamic diameter (AMAD).
Nineteen UF6/UO2F2 inhalation studies were undertaken in purebred, female beagle dogs (N = 16) to examine inter alia, (a) the possible relations of exposure, whole body, lung and renal uranium levels to excretion rates; (b) the threshold U6+ dose and renal concentration for renal injury; (c) the distribution and retention functions for U6+ in major tissues; (d) biochemical indicators of renal injury; and (e) aspects of U-induced tolerance. Each of these issues was investigated in the context of the chemical toxicity of U6+ following brief exposures to 235UO2F2 in the presence or absence of HF (the decomposition products of 235UF6). Both gamma-(235U) and alpha-(234U) counting methods were applied. In nine studies on 5 dogs, UO2F2 was administered intravenously. The major findings from both types of studies include: (1) UO2F2 retention time in the lungs is shorter than for UO3 or uranyl nitrate, viz. greater than 80% translocated with T 1/2 of less than 20 min; (2) the urinary elimination of U6+ follows closely to the ICRP excretion equation; (c) an absorbed dose of approximately 10 micrograms U6+ kg-1 body weight appears to be effective in producing renal injury; (d) a renal concentration of 0.3 micrograms g-1 kidney is close to a threshold concentration for renal injury; and (e) urinary and blood biochemical changes and histopathologic data were acquired and evaluated in both novice and tolerant animals. This report, considers all of these objectives and findings: Those involving biochemical indices and uranium-induced tolerance will be more fully reported elsewhere. In general, the dog studies attest to the usefulness of the intravenous human studies for certain U6+ dose-response data and interface well with new retention data on intravenous uranyl citrate in dogs by Stevens et al.
Dissolution characteristics of mixed-oxide nuclear fuels are important considerations for prediction of biological behavior of inhaled particles. Four representative industrial mixed-oxide powders were obtained from fuel fabrication enclosures. Studies of the dissolution of Pu, Am and U from aerosol particles of these materials in a serum simulant solution and in 0.1M HCl showed: (1) dissolution occurred at a rapid rate initially and slowed at longer times, (2) greater percentages of U dissolved than Pu or Am: with the dissolution rates of U and Pu generally reflecting the physical nature of the UO2-PuO2 matrix, (3) the temperature history of industrial mixed-oxides could not be reliably related to Pu dissolution except for a 3-5% increase when incorporated into a solid solution by sintering at 1750 degrees C, and (4) dissolution in the serum simulant agreed with the in vivo UO2 dissolution rate and suggested the dominant role of mechanical processes in PuO2 clearance from the lung. The rapid initial dissolution rate was shown to be related, in part, to an altered surface layer. The advantages and uses of in vitro solubility data for estimation of biological behavior of inhaled industrial mixed oxides, such as assessing the use of chelation therapy and interpretation of urinary excretion data, are discussed. It was concluded that in vitro solubility tests were useful, simple and easily applied to individual materials potentially inhaled by humans.
An attempt has been made to measure gamma radiation exposure rates from natural UO2 fuel pellets at near contact distances by means of a GM survey meter and LiF thermoluminescent dosimeters. An experimental device was fabricated which facilitated the holding of both the fuel pellets and the necessary measuring devices. A semi-empirical method for predicting exposure rates from a fuel pellet at near contact distances is also described. It is suggested that the experimental set up has the potential for providing a calibration device for gamma radiation instrumentation used to assess exposures in facilities handling U.
In this paper a comparison is made between the build-up of U thorax burdens and the predicted total lung (lung and lymph) burden, based on the lung model provided in ICRP Publication 30 for a group of 29 atomic radiation workers at a Canadian fuel fabrication facility. A similar comparison is made between the predicted ratio of the total lung burden to urinary excretion and the ratio obtained from bioassay data. The study period for the comparison is 5 y. The inhalation input for the lung model calculations was derived from air-sampling data and the choice of particle size activity median aerodynamic diameter (AMAD) was guided by particle size measurements made at representative work locations. The pulmonary clearance half-times studied were 100, 250 and 500 d. For the purpose of this comparison, averaged exposure and averaged bioassay data for the group were used. This comparison indicates that for the conditions of this facility, the assumption of a 500-d pulmonary clearance half-time and a particle size of 1 micron (AMAD) may be too conservative. It is suggested that measurements of air concentrations and particle size used as input parameters for the ICRP Publication 30 lung model may be used to calculate bioassay parameters which may then be tested against bioassay data obtained as part of an operational health physics program, thereby giving a useful step towards defining a derived air concentration value for U in the workplace.
Dissolution of gross alpha emitter radioactivity from particles composed of mixed uranium and plutonium oxides or of plutonium dioxide continually immersed in solvent typically display at least a two-phase dissolution pattern. Rapid dissolution of a small fraction of the total particulate mass is followed by much slower dissolution for the majority of the particulate mass. In this study, respirable particles of (U, Pu)O2 and PuO2 were subjected to dissolution using an alternate wetting and drying cycle. Particles were continuously immersed in solvent for 4 d and then dried in air for 3 d. This cycle was repeated weekly for 7 wk. Four solvents were used to represent a range of potential environmental conditions and a fifth solvent was used for comparison to other continuous immersion studies. In contrast to dissolution studies involving continuous immersion over periods of two or more weeks that exhibit a three-phase dissolution process, the alternate wet-dry cycling resulted in repetition of the first two phases of the dissolution pattern for each cycle. This led to significantly enhanced dissolution of both particulate materials. The enhancement in total dissolution ranged from two to ten times larger during each wet-dry cycle compared to studies involving continuous immersion. The results indicate a potential need to re-evaluate environmental models of actinide element bioavailability for particulate materials released to environments where wet-dry cycling may be routine, i.e. intermittent rainfall in an otherwise arid climate or in stream beds with intermittent flow.
This paper reports on an enriched uranium dioxide (UO2) mass clearance study undertaken with Fischer-344 rats. The UO2 had a uranium (U) isotopic composition of 0.79% 234U, 92.8% 235U, 0.34% 236U and 6.06% 238U, by mass, with an alpha-particle activity of 1.91 Bq micrograms-1. Forty-six rats were exposed to an enriched UO2 aerosol that had an activity median particle aerodynamic diameter ranging from 2.7 to 3.2 microns. The rats were killed from 1 to 720 d post-inhalation (PI). The mass of enriched UO2 present in the trachea, lung lobes, thoracic lymph nodes, kidneys, liver, spleen, gut, and the remainder of the carcass was assessed at death. At 720 d after exposure, 82% of the total body burden of enriched UO2 was in the lung, with a further 10% in the thoracic lymph nodes. This represented 17% and 2% of the original (5 d PI) lung burden. The mass clearance of enriched UO2 from the lung was fitted to a single exponential function, normalized to 100% at 5 d PI. The rate constant (k) was 2.8 X 10(-3) d-1, giving a clearance half-time of 247 d. Although statistical comparisons with the majority of published data were not possible, it appeared that both enriched UO2 and natural UO2 particles cleared at rates that were broadly similar, with a t1/2 in the rat lung of 150 to 300 d over the 5- to 720-d PI period. As a consequence of the 234U component in the inhaled UO2 particles, the rats killed at 720 d PI received a total mean accumulated alpha-particle dose averaged over the whole lung of 5.7 Gy. Histologic investigations of the rat lungs found that widespread lung disease was only observed in animals killed at 720 d PI.
Our experience with a U-intoxicated patient and human data in the literature suggest the kidney U burden should not be allowed to exceed 0.26 micrograms g-1. The Annual Limit on Intake (ALI) for natural U was calculated from the chemical toxicity of U and was found to be lower than that reported previously.
Quantifying radiological risk following the inhalation of radioactive aerosols entails not only an assessment of particle deposition within respiratory tract regions but a full accounting of clearance mechanisms whereby particles may be translocated to adjacent respiratory tissue regions, absorbed to blood, or released to the gastrointestinal tract. The model outlined in ICRP Publication 66 represents to date one of the most complete overall descriptions of particle deposition and clearance, as well as localized radiation dosimetry, within the respiratory tract. In this study, a previous review of the ICRP-66 deposition model is extended to the study of the subsequent clearance model. A systematic review of the clearance component within the ICRP 66 respiratory tract model was conducted in which probability density functions were assigned to all input parameters for both 239PuO2 and 238UO2/238U3O8. These distributions were subsequently incorporated within a computer code LUDUC (Lung Dose Uncertainty Code) in which Latin hypercube sampling techniques are used to generate multiple (e.g., 1,000) sets of input vectors (i.e., trials) for all model parameters needed to assess mechanical clearance and particle dissolution/absorption. Integral numbers of nuclear disintegrations, U(s), in various lung regions were shown to be well-described by lognormal probability distributions. Of the four extrathoracic clearance compartments of the respiratory tract, uncertainties in U(s), expressed as the ratio of its 95% to 5% confidence levels, were highest within the LN(ET) tissues for 239PuO2 (ratio of 50 to 130) and within the ET(seq) tissues for 238UO2/238U3O8 (ratio of 12 to 50). Peak uncertainties in U(s) in these respiratory regions occurred at particle sizes of approximately 0.5-0.6 microm where uncertainties in ET2 particle deposition fractions accounted for only approximately 10% of the total U(s) uncertainty for 239PuO2, and only approximately 30% of the total U(s) uncertainty for 238UO2/238U3O8 (the remainder is attributed to the clearance model alone). Of the eight clearance compartments within the thoracic regions of the respiratory tract, and for particle sizes below approximately 5 microm, uncertainties in U(s) were highest within the LN(TH) tissues for 239PuO2 (ratio of 60 to 80) and within the BB(seq) tissues for 238UO2/238U3O8 (ratio of 20 and 60). At particle sizes exceeding approximately 5 microm in aerodynamic diameter, peak uncertainties in U(s) are noted for the AI, bb(seq), and bb1 clearance compartments. As the particle size approaches 10 microm in size, uncertainties in U(s) within these three thoracic tissue regions approach a factor of 1,000 and are dominated by corresponding uncertainties in particle deposition.
This paper extends an examination of the influence of parameter uncertainties on regional doses to respiratory tract tissues for short-ranged alpha particles using the ICRP-66 respiratory tract model. Previous papers examined uncertainties in the deposition and clearance aspects of the model. The critical parameters examined in this study included target tissue depths, thicknesses, and masses, particularly within the thoracic or lung regions of the respiratory tract. Probability density functions were assigned for the parameters based on published data. The probabilistic computer code LUDUC (Lung Dose Uncertainty Code) was used to assess regional and total lung doses from inhaled aerosols of 239PuO2 and 238UO2/238U3O8. Dose uncertainty was noted to depend on the particle aerodynamic diameter. Additionally, dose distributions were found to follow a lognormal distribution pattern. For 239PuO2 and 238UO2/238U3O8, this study showed that the uncertainty in lung dose increases by factors of approximately 50 and approximately 70 for plutonium and uranium oxides, respectively, over the particle size range from 0.1 to 20 microm. For typical exposure scenarios involving both radionuclides, the ratio of the 95% dose fractile to the 5% dose fractile ranged from approximately 8-10 (corresponding to a geometric standard deviation, or GSD, of about 1.7-2) for particle diameters of 0.1 to 1 microm. This ratio increased to about 370 for plutonium oxide (GSD approximately 4.5) and to about 600 for uranium oxide (GSD approximately 5) as the particle diameter approached 20 microm. However, thoracic tissue doses were quite low at larger particle sizes because most of the deposition occurred in the extrathoracic airways. For 239PuO2, median doses from LUDUC were found be in general agreement with those for Reference Man (via deterministic LUDEP 2.0 calculations) in the particle range of 0.1 to 5 microm. However, median doses to the basal cell nuclei of the bronchial airways (BB(bas)) calculated by LUDUC were found to be approximately 6 times higher than LUDEP reference doses. The higher BB(bas) doses were directly attributed to discrepancies between the ICRP default thickness for the bronchial epithelium (55 microm) and the probability density function assumed within LUDUC (uniform distribution from 20 to 60 microm based upon detailed literature reviews).
The objective of this study was to establish a delayed air sampling method to verify the integrity of an existing HEPA filter system in a ventilated fume hood. (238U,232Th)O2 microspheres were generated to fabricate cement nuclear fuel pellets in a HEPA-filtered hood. To comply with the air effluent concentration limits by NRC, the capture efficiency of the HEPA filter was examined. An in-line isokinetic air sampling system was installed downstream of the HEPA filter. Utilizing a gas flow proportional counter, 212Pb was used as a surrogate to indicate any possible penetration of the (238U,232Th)O2 particles through the HEPA filter. Based on the experimental results, this delayed sampling method proved to be an easy and effective way to validate the integrity of the HEPA filter.
The ability of a naturally occurring Citrobacter sp. to accumulate cadmium has been attributed to cellular precipitation of CdHPO4, utilizing HPO4(2-) liberated via the activity of an overproduced, Cd-resistant acid-type phosphatase. Phosphatase production and heavy metal accumulation by batch cultures of this strain (N14) and a phosphatase-deficient mutant were compared with two reference strains of Citrobacter freundii. Only strain N14 expressed a high level of acid phosphatase and accumulated lanthanum and uranyl ion enzymically. Acid phosphatase is regulated via carbon-starvation; although the C. freundii strains overexpressed phosphatase activity in carbon-limiting continuous culture, this was approximately 20-fold less than the activity of strain N14 grown similarly. Citrobacter strain N14 was originally isolated from a metal-contaminated soil environment; phosphatase overproduction and metal accumulation were postulated as a detoxification mechanism. However, application of Cd-stress, and enrichment for Cd-resistant C. freundii ('training'), reduced the phosphatase activity of this organism by about 50% as compared to Cd-unstressed cultures. The acid phosphatase of C. freundii and Citrobacter N14 had a similar pattern of resistance to some diagnostic reagents. The enzyme of the latter is similar to the PhoN acid phosphatase of Salmonella typhimurium described by other workers; the results are discussed with respect to the known phosphatases of the enterobacteria.
A strain of Acinetobacter with potential for bioremediation of heavy metal-contaminated waters was isolated from a wastewater-treatment plant operating an enhanced biological phosphate removal process. NMR and extractive methods showed that polyphosphate accumulated aerobically was degraded under anaerobic conditions both in the presence and absence of cadmium or uranium (0.2-0.5 mM). NMR showed that free phosphate was formed at the expense of polyphosphate, and an extractive technique indicated that this reaction could be stimulated by the presence of UO(2)2+ under these conditions. Energy-dispersive X-ray microanalysis demonstrated that only cadmium could enter the cells, and co-localized with intra-cellular granules containing phosphate and other divalent metals. The effects of other environmental parameters on the anaerobic phosphate metabolism were also investigated. Between pH 5.5 and 8.0, phosphate release increased with increasing pH. Between 4 degrees C and 37 degrees C, phosphate release increased with increasing temperature. The presence of nitrate at concentrations of 10 mM and above inhibited anoxic phosphate release, but supplying tungstate in the growth medium prior to anoxic incubation reduced the production of active nitrate reductase and alleviated this effect.
Synchrotron experiments with uranium antiferromagnetic compounds have discovered large ( >1000) enhancements of the magnetic scattering intensities at the K edges of nominally nonmagnetic anions, e.g., Ga and As. The width in energy, the position with respect to the white line, and the azimuthal and polarization dependencies permit one to associate the signal with transitions of E1 dipole symmetry from 1s to 4p states. In momentum space, the signal exhibits long-range order at the antiferromagnetic wave vector. We discuss possible channels capable of generating the observed enhancements.