Anthony C. James

Elemental bio-imaging of thorium, uranium and plutonium in tissues from occupationally exposed former nuclear workers

Internal exposure from naturally occurring radionuclides (including the inhaled long-lived actinides (232)Th and (238)U) is a component of the ubiquitous background radiation dose (National Council on Radiation Protection and Measurements. Ionizing radiation exposure of the population of the United States; NCRP Report No. 160; NCRP: Bethesda, MD, 2009). It is of interest to compare the concentration distribution of these natural alpha-emitters in the lungs and respiratory lymph nodes with those resulting from occupational exposure, including exposure to anthropogenic plutonium and depleted and enriched uranium. This study examines the application of laser ablation-inductively coupled plasma-mass spectrometry (LA-ICPMS) to quantifying and visualizing the mass distribution of uranium and thorium isotopes from both occupational and natural background exposure in human respiratory tissues and, for the first time, extends this application to the direct imaging of plutonium isotopes. Sections of lymphatic and lung tissues taken from deceased former nuclear workers with a known history of occupational exposure to specific actinide elements (uranium, plutonium, or americium) were analyzed by LA-ICPMS. Using a previously developed LA-ICPMS protocol for elemental bio-imaging of trace elements in human tissue and a new software tool, we generated images of thorium ((232)Th), uranium ((235)U and (238)U), and plutonium ((239)Pu and (240)Pu) mass distributions in sections of tissue. We used a laboratory-produced matrix-matched standard to quantify the (232)Th, (235)U, and (238)U concentrations. The plutonium isotopes (239)Pu and (240)Pu were detected by LA-ICPMS in 65 mum diameter localized regions of both a paratracheal lymph node and a sample of lung tissue from a person who was occupationally exposed to refractory plutonium (plutonium dioxide). The average (overall) (239)Pu concentration in the lymph node was 39.2 ng/g, measured by high purity germanium (HPGe) gamma-spectrometry (Lynch, T. P.; Tolmachev, S. Y.; James, A. C. Radiat. Prot. Dosim. 2009, 134, 94-101). Localized mass concentrations of thorium ((232)Th) and uranium ((238)U) in lymph node tissue from a person not occupationally exposed to these elements (chronic natural background inhalation exposure) ranged up to 400 and 375 ng/g, respectively. In lung samples of occupationally nonexposed to thorium and uranium workers, (232)Th and (238)U concentrations ranged up to 200 and 170 ng/g, respectively. In a person occupationally exposed to air-oxidized uranium metal (Adley, F. E.; Gill, W. E.; Scott, R. H. Study of atmospheric contaminiation in the melt plant buiding. HW-23352(Rev.); United States Atomic Energy Commission: Oakridge, TN, 1952, p 1-97), the maximum (235)U and (238)U isotopic mass concentrations in a lymph node, measured at higher resolution (with a 30 mum laser spot diameter), were 70 and 8500 ng/g, respectively. The ratio of these simultaneously measured mass concentrations signifies natural uranium. The current technique was not sufficiently sensitive, even with a 65 mum laser spot diameter, to detect (241)Am (at an overall tissue concentration of 0.024 ng/g, i.e., 3 Bq/g).

Assessment of the exposure to and dose from radon decay products in normally occupied homes

The exposure to radon decay products has been assessed in seven homes in the northeastern United States and southeastern Canada. In two of the houses, there was a single individual who smoked cigarettes. There were a variety of heating and cooking appliances among these homes. These studies have provide 565 measurements of the activity-weighted size distributions in these houses. The median value for the equilibrium factor was 0.408 as compared with the previously employed value of 0.50. Using the recently adopted ICRP lung deposition and dosimetry model, the hourly equivalent lung dose rate per unit, radon exposure was estimated for each measured size distribution. The mean equivalent dose rate per unit of 222 Rn gas concentration was approximately 140 nSv h -1 Bq -1 m -3 . It was found that the equivalent dose was strongly correlated with the ratio of the decay product concentration to that of radon, termed the equilibrium factor, F, with a correlation coefficient of 0.785. The correlation coefficient with the ≤2-nm size fraction (the «unattached» fraction) was 0.169, reflecting no significant relationship with the unattached fraction. Differences between houses with smokers present and absent were noted in the exposure conditions, but the resulting dose rate per unit of radon gas concentration was essentially the same for the two groups. Expressed in terms of ICRPs unit of effective dose for members of the public, the mean dose rate conversion coefficient with respect to radon gas concentration found in this study was 3.8 nSv h -1 Bq -1 m -3

Unattached fraction measuring technique and radon lung dose

We examined the influence of measurement techniques used to characterize the ultrafine fraction of airborne radon progeny on the estimation of dose rate received by the lungs by comparing simultaneous measurements of the unattached fraction and the full activity-weighted size distribution. The unattached fraction was measured using a single stainless steel-wire-mesh screen (50% penetration for 4 nm equivalent diameter particles). The full activity-weighted size distribution was measured using a graded-screen array (six stages with equivalent particle diameter for 50% penetration ranging from 0.5 to 500 nm). Thirty paired samples were taken outdoors in a rural location in central New Mexico, USA. Using current exposure-dose conversion factors based on the ICRP 66 Lung Model, we compared the calculated effective dose rate for the respiratory tract for each sample pair. On average, the dose calculated from the full radon-progeny particle-size distribution (graded-screen array measurements) was 80% higher than that calculated from the single-screen measurements. Our results raise questions about the accuracy (i.e., potential low bias) of retrospective dose estimates for miners and other radon-exposed groups for whom the dose assessment was based on the nominal unattached fraction measured with a single screen.

Microdistribution and Long-term Retention of 239Pu (NO3)4 in the Respiratory Tracts of an Acutely Exposed Plutonium Worker and Experimental Beagle Dogs

The long-term retention of inhaled soluble forms of plutonium raises concerns as to the potential health effects in persons working in nuclear energy or the nuclear weapons program. The distributions of long-term retained inhaled plutonium-nitrate [(239)Pu (NO(3))(4)] deposited in the lungs of an accidentally exposed nuclear worker (Human Case 0269) and in the lungs of experimentally exposed beagle dogs with varying initial lung depositions were determined via autoradiographs of selected histologic lung, lymph node, trachea, and nasal turbinate tissue sections. These studies showed that both the human and dogs had a nonuniform distribution of plutonium throughout the lung tissue. Fibrotic scar tissue effectively encapsulated a portion of the plutonium and prevented its clearance from the body or translocation to other tissues and diminished dose to organ parenchyma. Alpha radiation activity from deposited plutonium in Human Case 0269 was observed primarily along the subpleural regions while no alpha activity was seen in the tracheobronchial lymph nodes of this individual. However, relatively high activity levels in the tracheobronchial lymph nodes of the beagles indicated the lymphatic system was effective in clearing deposited plutonium from the lung tissues. In both the human case and beagle dogs, the appearance of retained plutonium within the respiratory tract was inconsistent with current biokinetic models of clearance for soluble forms of plutonium. Bound plutonium can have a marked effect on the dose to the lungs and subsequent radiation exposure has the potential to increase cancer risk.

USTUR whole body case 0262: 33-y follow-up of Puo2 in a skin wound and associated axillary node

This whole body donation case (USTUR Registrant) involved two suspected PuO2 inhalation intakes, each indicated by a measurable Pu alpha activity in a single urine sample, followed about 1(1/2) y later by a puncture wound to the thumb while working in a Pu glovebox. The study is concerned with modelling simultaneously the biokinetics of deposition and retention in the respiratory tract and at the wound site; and the biokinetics of Pu subsequently transferred to other body organs, until the donors death. Urine samples taken after the wound incident had readily measurable Pu alpha activity over the next 14 y, before dropping below the minimum detectable excretion rate (<0.4 mBq d(-1)). The Registrant died about 33 y after the wound intake, at the age of 71, from hepatocellular carcinoma with extensive metastases. At autopsy, all major soft tissue organs were harvested for analysis of their 238Pu, 239+240Pu and 241Am content. The amount of 239+240Pu retained at the wound site was 68 +/- 7 Bq (1 SD), measured by low-energy planar Ge spectrometry. A further 56.0 +/- 1.2 Bq was retained in an associated axillary lymph node, measured by radiochemistry. Simultaneous mathematical analysis (modelling) of all in vivo urinary excretion data, together with the measured lung, thoracic lymph node, wound, axillary lymph node and systemic tissue contents at death, yielded estimated intake amounts of 757 and 1504 Bq, respectively, for the first and second inhalation incidents, and 204 Bq for the total wound intake. The inhaled Pu material was highly insoluble, with an estimated long-term absorption rate from the lungs of 2 x 10(-5) d(-1). The Pu material deposited at the wound site was mixed: approximately 14% was rapidly absorbed, approximately 49% was absorbed at the rate of about 6 x 10(-5) d(-1), and the remainder ( approximately 37%) was absorbed extremely slowly (at the rate of about 5 x 10(-6) d(-1)). Thus, it was estimated that only approximately 40% of the Pu initially deposited in the wound had been absorbed systemically over the 33-y period until the donors death. The biokinetic modelling also indicated that, in this individual case, some of the parameter values (rate constants) incorporated in the ICRP Publication 67 Pu model were up to a factor of 2 different from ICRPs recommended values (for reference man).

Reevaluation of USTUR plutonium wound case 0262 using Bayesian methodology and new data.

Abstract Skin penetration by radionuclide contaminants serves as a route of entry into the body and may pose a serious health risk to humans depending on the magnitude of intake. The United States Transuranium and Uranium Registry whole body Case 0262 was involved in a wound intake of plutonium at the Hanford Site. The registrant died about 33 years later. Results were initially reported in 2007 regarding the deposition and retention of plutonium in various tissues, including the wound site. However in 2009, an additional (previously unrecorded) sample of the wound tissue was located in the National Human Radiobiological Tissue Repository. The new sample was analyzed using inductively coupled plasma-mass spectrometry (ICP-MS), and the results were used to calibrate the measurement of emitted 239Pu x-rays from the original wound tissue sample made in 2007. In the present study, the analysis of 239Pu absorption rates from the wound and axillary lymph node from the initial study is repeated using the additional wound activity data and ICP-MS calibration. This new analysis is carried out using the Weighted Likelihood Monte Carlo Sampling (WeLMoS) method and code, which applies Bayesian inference to calculate the posterior probability distribution of intake and wound absorption parameters directly from the observed data and the assumed biokinetic model structure. The resulting central estimates of empirical wound absorption parameters and their associated uncertainties are here compared with the empirical values recommended in NCRP Report No. 156 for plutonium and with the maximum likelihood point estimates derived in the initial study from the Case 0262 data available at the time.

Experimental and modeling studies of 220Rn decay products in outdoor air near the ground surface

Measurements of the concentrations of 220Rn gas and its short-lived decay products (216Po, 212Pb, 212Bi) in attached-to-aerosol and unattached states were carried out at a semi-arid field site in central New Mexico under varying atmospheric conditions. A high volume air sampler (100 L min−1), with a single 635 mesh stainless steel screen (50% penetration for 4 nm particles) and a glass fiber filter, were employed. Analysis of the 105 outdoor measurements yielded the following average values: 9 Bq m−3 for the 220Rn activity concentration; 22 nJ m−3 for the concentration of potential alpha energy, 0.03 for the equilibrium factor, 0.15 for the unattached fraction of potential alpha energy, and 14 nSv h−1 for the effective dose rate. The measurements were interpreted using the computer code TPOUT, which applies one-dimensional eddy transport to calculate the vertical concentrations of 220Rn and its decay products as a function of atmospheric stability, aerosol concentration, terrain roughness, and surface wind speed. This code successfully predicted the observed trends of unattached and attached decay products concentrations with height, and demonstrated the importance of the mixing height and the rate of vertical transport for controlling breathing level dose. Projection of the present results to conditions typical of temperate climates suggests an average annual outdoor effective dose from 220Rn decay products of about 0.025 mSv, which is a significant component of the total outdoor dose.

Maximum Likelihood Analysis of Bioassay Data from Long-term Follow-up of Two Refractory Puo2 Inhalation Cases

AbstractThe U.S. Transuranium and Uranium Registries’ tissue donors 0202 and 0407 are the two most highly exposed of the 18 registrants who were involved in the 1965 plutonium fire accident at a defense nuclear facility. Material released during the fire was well characterized as “high fired” refractory plutonium dioxide with 0.32-&mgr;m mass median diameter. The extensive bioassay data from long-term follow-up of these two cases were used to evaluate the applicability of the Human Respiratory Tract Model presented by International Commission on Radiological Protection in Publication 66 and its revision proposed by Gregoratto et al. in order to account for the observed long-term retention of insoluble material in the lungs. The maximum likelihood method was used to calculate the point estimates of intake and tissue doses and to examine the effect of different lung clearance, blood absorption, and systemic models on the goodness-of-fit and estimated dose values. With appropriate adjustments, Gregoratto et al. particle transport model coupled with the customized blood absorption parameters yielded a credible fit to the bioassay data for both cases and predicted the Case 0202 liver and skeletal activities measured postmortem. PuO2 particles produced by the plutonium fire are extremely insoluble. About 1% of this material is absorbed from the respiratory tract relatively rapidly, at a rate of about 1 to 2 d−1 (half-time about 8 to 16 h). The remainder (99%) is absorbed extremely slowly, at a rate of about 5 × 10−6 d−1 (half-time about 400 y). When considering this situation, it appears that doses to other body organs are negligible in comparison to those to tissues of the respiratory tract. About 96% of the total committed weighted dose equivalent is contributed by the lungs. Doses absorbed by these workers’ lungs were high: 3.2 Gy to AI and 6.5 Gy to LNTH for Case 0202 (18 y post-intake) and 3.2 Gy to AI and 55.5 Gy to LNTH for Case 0407 (43 y post-intake). This evaluation supports the Gregoratto et al. proposed revision to the ICRP 66 model when considering situations of extremely insoluble particles.

The US Transuranium and Uranium Registries: forty years' experience and new directions in the analysis of actinides in human tissues

Abstract The US Transuranium and Uranium Registries (USTUR) studies the distribution, biokinetics and tissue dosimetry of actinide elements through radiochemical analysis of autopsy tissues voluntarily donated by occupationally exposed persons. The paper provides an overview of the analytical methods for plutonium (Pu), americium (Am) and uranium (U) isotopic determination in human tissues currently applied at USTUR. The results of inter-comparing 239+240Pu, 241Am and 234,235,238U determinations by sector field inductively coupled mass spectrometry (SF-ICP-MS), α-spectrometry (AS) and kinetic phosphorescence analysis (KPA) are discussed. SF-ICP-MS is a major advance over AS and KPA in enabling the measurement of the 240Pu/ 239Pu atom ratio, the short-lived β-emitter 241Pu, and long-lived 236U. For the first time, 241Am and 241Pu were measured in human tissues using SF-ICP-MS. The paper also presents a new avenue of USTUR research in the application of laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) to elemental bio-imaging (EBI) of the actinides in human tissues.

Estimating 241Am activity in the body: comparison of direct measurements and radiochemical analyses

The assessment of dose and ultimately the health risk from intakes of radioactive materials begins with estimating the amount actually taken into the body. An accurate estimate provides the basis to best assess the distribution in the body, the resulting dose and ultimately the health risk. This study continues the time-honoured practice of evaluating the accuracy of results obtained using in vivo measurement methods and techniques. Results from the radiochemical analyses of the (241)Am activity content of tissues and organs from four donors to the United States Transuranium and Uranium Registries (USTUR) were compared with the results from direct measurements of radioactive material in the body performed in vivo and post-mortem. Two were whole-body donations and two were partial-body donations. The (241)Am lung activity estimates ranged from 1 to 30 Bq in the four cases. The (241)Am activity in the lungs determined from the direct measurements were within 40% of the radiochemistry results in three cases and within a factor of 2 for the other case. However, in one case the post-mortem direct measurement estimate was a factor of 10 higher than the radiochemistry result for lung activity, most probably due to underestimating the skeletal contribution to the measured count rate over the lungs. The direct measurement estimates of liver activity ranged from 2 to 60 Bq and were consistently lower than the radiochemistry results. The skeleton was the organ with the highest deposition of (241)Am activity in all four cases. The skeletal activity estimates ranged from 30 to 300 Bq. The skeletal activity obtained from measurements over the forehead were within 20% of the radiochemistry results in three cases and differed by 78% in the other case. The results from this study suggest that the measurement methods, data analysis methods and calibration techniques used at the In Vivo Radiobioassay and Research Facility can be used to quantify the activity in the lungs, skeleton and liver when (241)Am activity is present in all three organs. The adjustment method used to account for the contribution from activity in other organs improved the agreement between the direct measurement results and the radiochemistry results for activity in the lungs and skeleton. The method appeared to overestimate the contribution from the other organs to the liver activity measurements, although the low activity levels complicated the analysis. The unadjusted liver activity estimates from the direct measurements were generally in better agreement with the radiochemistry results than the adjusted liver activity. The data from this study indicates that the results from the in vivo measurement techniques provide reasonable estimates of radioactive material in the lungs and skeleton under the most challenging conditions where there is (241)Am activity in multiple organs. The data analysis from additional USTUR cases with both direct measurement results and radiochemistry results is in progress to further evaluate how best to account for the contributions from (241)Am activity in multiple organs and to better understand the uncertainty associated with the adjusted activity.