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Bioassay of radiation workers handling actinides in powder form is carried out annually under routine monitoring program. In case of routine monitoring, as the day of intake is unknown, it is assumed that intake has occurred at the midpoint of monitoring interval. However, an uncertainty is introduced in intake estimation due to this assumption. In the present work, uncertainties were evaluated due to unknown time of intake of Type M and S compounds of Pu and U for annual and biennial monitoring interval. Lung and excreta monitoring methods were considered for uncertainty estimation along with the spread of ±1 month or ±2 months in measurements, at the end of monitoring interval. Annual lung monitoring is found to be the most robust and accurate method for assessment of intake of Type S compounds of Pu, Am and U when time of intake is unknown and Pu is detected using 241 Am as a tracer. For Type M compounds, uncertainty was found to be higher than 3 for biennial monitoring interval. Uncertainty is found to be less than 3 for Type S compounds of Pu and U and also for Type M compounds of Pu for both, annual and biennial monitoring intervals using urine analysis. It is higher than 3 for Type M compounds of U for both the monitoring intervals. It is greater than 3 for both Type M and S compounds of Pu and U for both the monitoring intervals using fecal analysis., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 Elsevier Ltd. All rights reserved.)

Abstract: In the nuclear industry, wound contamination with americium is expected to increase with decommissioning and waste management. Treatment of workers with diethylenetriaminepentaacetic acid (DTPA) requires optimization to reduce internal contamination and radiation exposure. This work aimed at evaluating and comparing different DTPA protocol efficacies after wound contamination of rats with americium. Wound contamination was simulated in rats by depositing americium nitrate in an incision in the hind limb. Different routes, times, and frequencies of DTPA administration were evaluated. Individual daily urinary americium excretion and tissue retention were analyzed using the statistical tool STATBIODIS. Urinary profiles, urinary enhancement factors, and inhibition percentages of tissue retention were calculated. A single DTPA administration the day of contamination induced a rapid increase in americium urinary excretion that decreased exponentially over 7 d, indicating that the first DTPA administration should be delivered as early as possible. DTPA treatment limited americium uptake in systemic tissues irrespective of the protocol. Liver and skeleton burdens were markedly reduced, which would drive reduction of radiation dose. Local or intravenous injections were equally effective. Inherent difficulties in wound site activity measurements did not allow identification of a significant decorporating effect at the wound site. Repeated intravenous injections of DTPA also increased americium urinary excretion, which supports the use of multiple DTPA administrations shortly after wound contamination. Results from these statistical analyses will contribute to a better understanding of americium behavior in the presence or absence of DTPA and may aid optimization of treatment for workers., Competing Interests: The authors declare no conflicts of interest., (Copyright © 2021 Health Physics Society.)

The implementation of the one-pass-through separation technique using two stacked chromatography columns of TEVA - TRU resins for the separation of 237 Np, 241 Am, thorium, plutonium and uranium from environmental and urine samples was investigated. The sequential separation technique proved to be successful and gave similar results to those obtained when using individual separations. The analysis time was considerably improved. The amount of chemical waste was also reduced by 50% and the use of HClO 4 was avoided. The technique of ICP-MS was also investigated as a complementary technique to alpha-spectrometry., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

An americium solution injected intramuscularly into several nonhuman primates (NHPs) was found to behave differently than predicted by the wound models described in the NCRP Report 156. This was because the injection was made along with a citrate solution, which is known to be more soluble than chlorides, oxides, or nitrates on which the NCRP Report was based. A multi-exponential wound model specific to the injected americium solution was developed based on the retention in the intramuscular sites. The model was coupled with the americium systemic model to interpret the urinary excretion data and assess the intake, and it was determined that the models were adequate to predict early urinary excretion in most cases but unable to predict late urinary excretion. This was attributed to the differences in the systemic handling of americium between humans and nonhuman primates. Information on the type of wounds, solubility, particle size, mass, chemical form, etc., should always be considered when performing wound dosimetry.

An intercomparison exercise on the determination of (241)Am, (244)Cm and (252)Cf in urine was performed. Since it was designed with regard to emergency preparedness, the detection limit for each nuclide was set to 0.1 Bq per 24-h urine sample. Most of the participating laboratories were established bioassay laboratories. However, some laboratories that routinely determine (241)Am only in environmental samples were also invited in order to explore their potential for emergency bioassay analysis. Another aspect of the intercomparison was to investigate the performance of all laboratories concerning the chemical yields of the (243)Am tracer in comparison with (244)Cm and (252)Cf. In summary, both types of laboratories showed good results. There was a negative bias for the results of (244)Cm and (252)Cf, which can be explained by slightly different radiochemical behaviours of americium, curium and californium and which is in agreement with results reported in the literature., (© The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.)

Two essential technical requirements of ISO 17025 guide for accreditation of testing and calibration laboratories are the validation of methods and the estimation of all sources of uncertainty that may affect the analytical result. Bioelimination Laboratory from Radiation Dosimetry Service of CIEMAT (Spain) uses alpha spectrometry to quantify alpha emitters (Pu, Am, Th, U and Cm isotopes) in urine and faecal samples from workers exposed to internal radiation. Therefore and as a step previous to achieving the ISO 17025 accreditation, the laboratory has performed retrospective studies based on the obtained results in the past few years to validate the analytical method. Uncertainty estimation was done identifying and quantifying all the contributions, and finally the overall combined standard uncertainty was calculated., (© The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.)

In case of internal contamination with plutonium materials, a treatment with diethylene triamine pentaacetic acid (DTPA) can be administered in order to reduce plutonium body burden and consequently avoid some radiation dose. DTPA intravenous injections or inhalation can start almost immediately after intake, in parallel with urinary and fecal bioassay sampling for dosimetric follow-up. However, urine and feces excretion will be significantly enhanced by the DTPA treatment. As internal dose is calculated from bioassay results, the DTPA effect on excretion has to be taken into account. A common method to correct bioassay data is to divide it by a factor representing the excretion enhancement under DTPA treatment by intravenous injection. Its value may be based on a nominal reference or observed after a break in the treatment. The aim of this study was to estimate the influence of this factor on internal dose by comparing the dose estimated using default or upper and lower values of the enhancement factor for 11 contamination cases. The observed upper and lower values of the enhancement factor were 18.7 and 63.0 for plutonium and 24.9 and 28.8 for americium. For americium, a default factor of 25 is proposed. This work demonstrates that the use of a default DTPA enhancement factor allows the determination of the magnitude of the contamination because dose estimated could vary by a factor of 2 depending on the value of the individual DTPA enhancement factor. In case of significant intake, an individual enhancement factor should be determined to obtain a more reliable dose assessment.

Ultralow level analysis of actinides in urine samples may be required for dose assessment in the event of internal exposures to these radionuclides at nuclear facilities and nuclear power plants. A new bioassay method for analysis of sub-femtogram levels of Am and Cm in large-volume urine samples was developed. Americium and curium were co-precipitated with hydrous titanium oxide from the urine matrix and purified by column chromatography separation. After target preparation using mixed titanium/iron oxides, the final sample was measured by compact accelerator mass spectrometry. Urine samples spiked with known quantities of Am and Cm isotopes in the range of attogram to femtogram levels were measured for method evaluation. The results are in good agreement with the expected values, demonstrating the feasibility of compact accelerator mass spectrometry (AMS) for the determination of minor actinides at the levels of attogram/liter in urine samples to meet stringent sensitivity requirements for internal dosimetry assessment.

Unlabelled: Abstract Purpose: To compare data on the whole-body distribution of americium-241 ((241)Am) in rats following intravenous injection (IV), inhalation, and wound (intramuscular injection, IM)., Material and Methods: Following exposure, each rat was placed in an individual metabolism cages for the duration of the study, 28 days (d). Urine and feces were collected daily. Tissues and organs were collected and measured., Results: Liver and skeleton were the main sites of deposition for all routes of exposure but the content differed substantially. By 28 d, (241)Am content in liver was similar for IV and IM administrations (12 ± 4% and 14 ± 5%, respectively), which was 3-fold higher compared to inhalation. Americium-241 content in skeleton was 27% by the end of the IV study; which was 50% higher compared to the IM study and 6-fold higher compared to inhalation. The cumulative excretion in 28 d was 54% for IV (44% by feces and 10% by urine); 38% for IM (34% by feces and 4% by urine); and 84% for inhalation (83% by feces and 1% by urine)., Conclusion: Unperturbed rat models for the three routes of administration are the baseline for evaluating the efficacy of chelating agents.

An internal contamination with (241)Am was detected in a worker during a routine monitoring of workers from a company producing Am sources for smoke detectors and Am-Be neutron sources. During the 4-year period after the exposure, the number of urine and faecal samples from the worker were analysed; in vivo measurements were also performed. Specific values for absorption parameters of the human respiratory tract model and particle transport values were applied to improve the model fit to the measurement data. A good agreement of the bioassay data with the so-modified model predictions was obtained.

Three workers incurred inhalation exposures to Am oxide as a result of waste sorting and compaction activities. The exposure magnitudes were not fully recognized until the following day when an in-vivo lung count identified a significant lung deposition of Am in a male worker, and DTPA chelation therapy was initiated. Two additional workers (one female and one male) were then identified as sufficiently exposed to also warrant therapy. In-vivo bioassay measurements were performed over the ensuing 6 mo to quantify the 241Am activity in the lungs, liver, and skeleton. Urine and fecal samples were collected and showed readily detectable 241Am. Clinical lab tests and medical evaluations all showed normal results. There were no significant adverse clinical health effects from the therapy. The estimated 241Am inhalation intakes for the three workers were 1,800 Bq, 630 Bq, and 150 Bq. Lung retention showed somewhat longer pulmonary clearance half-times than standard inhalation class W or absorption Type M assumptions. The three subjects underwent slightly different therapy regimens, with therapy effectiveness factors (defined as the ratio of the reference doses without therapy relative to the final assessed doses) of 4.5, 1.9, and 1.7, respectively.

The aim of this study is to propose a single modeling structure to describe both plutonium and americium decorporation by DTPA, which is based on hypotheses mostly validated by experimental data. Decorporation efficacy of extracellular retention depends on the concentration ratio of DTPA vs. actinides and varies in each compartment according to the amount of biological ligands and their affinity for actinides. By contrast, because the relatively long residence time of DTPA after its cell internalization and the stability of actinide-DTPA complexes, intracellular decorporation efficacy is mainly controlled by a DTPA/actinide ratio, which is specific to each retention compartment. Although the affinity of DTPA is much lower for americium than for plutonium, a larger decorporation of americium can be obtained, which is explained by different biological ligands and/or their affinity for the actinide. Altogether, these results show that the relative contribution of intra vs. extracellular decorporation varies depending on the actinide, the chemical form of radionuclides, the galenic formulation of DTPA, and the treatment schedule.

A 1985 plutonium puncture wound resulted in the initial deposition of 48 kBq of transuranic alpha activity, primarily 239+240Pu and 241Am, in a worker's right index finger. Surgical excisions in the week following reduced the long-term residual wound activity to 5.4 kBq, and 164 DTPA chelation therapy administrations over 17 mo resulted in urinary excretion of about 7 kBq. The case was published in 1988, but now 24 y of follow-up data are available. Annual bioassays have included in-vivo measurements of 241Am in the wound, skeleton, liver, lung, and axillary lymph nodes, and urinalyses for plutonium and 241Am. These measurements have shown relatively stable levels of 241Am at the wound site, with gradually increasing amounts of 241Am detected in the skeleton. Liver measurements have shown erratic detection of 241Am, and the lung measurements indicate Am but as interference from activity in the axillary lymph nodes and skeleton rather than activity in the lung. Urine excretion of Pu since termination of chelation therapy has typically ranged from 10 to 20 mBq d, with Am excretion about 10% of that for 239+240Pu. Annual routine medical exams have not identified any adverse health effects associated with the intake.

A group of workers internally contaminated with Am have been followed for about 12 years. The source of contamination was AmO2 powder used for production of AmBe neutron sources and other applications. The production of some radionuclide sources included chemical treatment of the original material, which transformed the americium into the nitrate, but mostly powder metallurgy was used for production of sources for smoke detectors. In vivo measurement of the workers was performed with two LEGe detectors placed near the head of the measured person. Calibration was performed with four different physical skull phantoms of different origin and a voxel phantom with Monte Carlo simulation, which was developed to fit the head sizes of individual persons. Samples of urine and feces were analyzed by means of radiochemical separation followed by alpha-spectrometry. Separation of 241Am from mineralized excreta was performed by combined anion exchange and extraction chromatographic techniques. As a tracer, 243Am was used. When the measured data (83 data on skeletal activity, activity in 389 bioassay samples) were compared with International Commission on Radiological Protection's and Leggett's biokinetic models of americium, it was found that in most cases, after more than 15 y since the intake, the excretion rate was lower (or skeletal activity higher) than predicted. On the other hand, the ratio of excreted activity in urine and feces agrees well with model predictions.

An emergency urine bioassay method has been developed for the determination of (241)Am in human urine samples. The method is based on extraction chromatographic separation of (241)Am from urine on a single DGA (N,N,N',N'-tetraoctyldiglycolamide) resin column followed by liquid scintillation counting of (241)Am. The minimum detectable activity (MDA) for the method was 0.02 Bq. Considering the volume of urine sample (17.2 ml) used by the method; the MDA was 1.3 Bq l(-1). Measurement accuracy (relative bias, B(r)) and repeatability (relative precision, S(B)) of the method were found to be -3.4 and 8.9 %, respectively, when urine samples were spiked with (241)Am (20 Bq l(-1)). Excellent linearity (r(2) > 0.999) was established over the range of 2-200 Bq l(-1). The method was also found to be robust (S(B)=10.2 %) against matrix effects from different urine samples. Performance of the rapid bioassay method for accuracy and repeatability were evaluated against the performance criteria for radiobioassay (ANSI N13.30) and found to be in compliance. Considering the simplicity, excellent analytical figures of merit and fast sample turnaround time (<1 h), it is a very promising rapid bioassay method for supporting the medical response to an emergency where internal contamination of (241)Am is involved.

The selectivity of a rapid (90)Sr bioassay technique over (241)Am, (238/239)Pu, (210)Po, (137)Cs and (60)Co has been investigated. Similar to (90)Sr, these radionuclides are likely to be used in radiological dispersive devices. The purpose of this study was to demonstrate the degree to which the (90)Sr bioassay technique is free from interference by these radionuclides if present in a urine matrix. The interfering radionuclides were removed (from (90)Sr) by their retention on an anion exchange column. While, recovery of the target radionuclide ((90)Sr) was found to be >or= 90 %, contributions from (241)Am, (242)Pu and (208)Po were found to be

A new rapid separation method that allows separation and preconcentration of actinides in urine samples was developed for the measurement of longer lived actinides by inductively coupled plasma mass spectrometry (ICP-MS) and short-lived actinides by alpha spectrometry; a hybrid approach. This method uses stacked extraction chromatography cartridges and vacuum box technology to facilitate rapid separations. Preconcentration, if required, is performed using a streamlined calcium phosphate precipitation. Similar technology has been applied to separate actinides prior to measurement by alpha spectrometry, but this new method has been developed with elution reagents now compatible with ICP-MS as well. Purified solutions are split between ICP-MS and alpha spectrometry so that long- and short-lived actinide isotopes can be measured successfully. The method allows for simultaneous extraction of 24 samples (including QC samples) in less than 3h. Simultaneous sample preparation can offer significant time savings over sequential sample preparation. For example, sequential sample preparation of 24 samples taking just 15 min each requires 6h to complete. The simplicity and speed of this new method makes it attractive for radiological emergency response. If preconcentration is applied, the method is applicable to larger sample aliquots for occupational exposures as well. The chemical recoveries are typically greater than 90%, in contrast to other reported methods using flow injection separation techniques for urine samples where plutonium yields were 70-80%. This method allows measurement of both long-lived and short-lived actinide isotopes. (239)Pu, (242)Pu, (237)Np, (243)Am, (234)U, (235)U and (238)U were measured by ICP-MS, while (236)Pu, (238)Pu, (239)Pu, (241)Am, (243)Am and (244)Cm were measured by alpha spectrometry. The method can also be adapted so that the separation of uranium isotopes for assay is not required, if uranium assay by direct dilution of the urine sample is preferred instead. Multiple vacuum box locations may be set-up to supply several ICP-MS units with purified sample fractions such that a high sample throughput may be achieved, while still allowing for rapid measurement of short-lived actinides by alpha spectrometry.

The Human Monitoring Laboratory has extended the use of its portable whole body counters to portable gamma spectrometers for urinalysis. The protocol tested measured a 120-mL sample in a polypropylene sample container for 5 min. Minimum detectable activities were estimated for 241Am, 57Co, 137Cs, and 60Co. The former is 113 Bq per sample, and the latter three are between 27-29 Bq per sample. Assuming an intake 5 d before the measurement, and all other parameters as default, the committed effective doses are 517 Sv, 76 muSv, 402 muSv, and 1.5 mSv, respectively. Clearly, this instrument can be used as a field deployable gamma spectrometer for urinalysis for activation and fission products, but actinides (and other low energy photon emitters) remain problematic.

A 38-y-old Caucasian male who suffered an acute accidental inhalation intake of 6.3 kBq of 241Am was monitored over 2,135 d using periodic in vivo measurements of the activity in the lungs, liver, and skeleton. Lung clearance was described by a two-compartment exponential model with half-times of 110 d and 10,000 d. The observed uptake of 241Am in the liver (72 Bq) and skeleton (170 Bq) was significantly greater than predicted by the ICRP models for liver (5 Bq) and skeleton (8 Bq). The half-time in the liver was approximately 850 d. Estimates of skeletal activity based on head, wrist, and knee counts generally agreed within 25% over the course of the monitoring period. The half-time in the skeleton was approximately 20,000 d.

A group of workers with occupational intakes of 241Am, which occurred a long time ago, has been followed for some time. Results of in vivo measurement and bioassay of excreta are compared with the values predicted by the ICRP Publication 78 model. The observed skeletal content is, as a rule, higher than the predicted one. The ratio of excreted activity in urine to that in faeces is in very good agreement with the model prediction. Another group of workers from a waste management department, who were internally contaminated in July 2001, has also been followed. In some cases, there is quite a large difference in calculated intake between excretion by urine and that by faeces. The contaminant was presumably the same as that in the group of workers with old intakes, but its physical and chemical form could be influenced by a fixating lacquer used to prevent the spread of contamination.

A worker inhaled 241AmO2 powder. Air sampling showed low activities but a nose blow revealed 92 Bq. Results from faecal sampling and lung and whole-body monitoring indicated an intake of about 200 Bq, but urine sampling, though commencing only 1 d after intake, showed below-threshold activities (< 0.2 mBq). This conflicts with predictions based on the ICRP Publication 67 biokinetic model for americium and the ICRP Publication 66 model for the human respiratory tract, if default lung parameters are used.

A worker noted a small wound to his thumb when leaving a work site that was undergoing decontamination because of past operations with plutonium (Pu) and americium (Am). Direct surveys of the wound site confirmed the presence of contamination. The chelating agent Ca-DTPA was administered via a nebuliser within an hour after discovery of the wound. External measurements were made of the wound site and wound dressings; 24-h urinary excretion data were collected periodically and the Pu and Am urine content was determined. Zn-DTPA was administered on three occasions. The ICRP Pu systemic model was modified to consider the enhanced urinary excretion following administration of the chelating agents. The analysis indicated that the wound resulted in an initial deposition of 400 Bq 238Pu, 2240 Bq (239/240)Pu and 1060 Bq 241Am. About 70% of the initial wound activity was removed by surgical procedures and less than 1% of the wound activity was removed by chelation therapy. This paper compares the observed urinary excretion data with that indicated by a simulation of the kinetics of the transfer from the wound site and the kinetics of the chelating agent and Pu.

Two well characterised Pu inhalation cases show some remarkable similarities between substantially different types of Pu oxide. The circumstances of exposure, therapy, bioassay data, chemical solubility studies and dosimetry associated with these cases suggest that highly insoluble Pu may be more common than previously thought, and can pose significant challenges to bioassay programmes.

The determination of actinides concentration level in excreta, mainly urine is currently carried out to monitor people potentially exposed to alpha emitters. To measure actinides in such samples, specific analytical protocols have been set up. The chemical purification uses different chromatographic columns to selectively separate the actinides and each fraction, after electroplating, is measured by alpha spectrometry. To reach 1 mBq l(-1) of U, Pu or Am using these protocols, 6 days equally distributed between the chemical purification and the measurement are necessary. The protocol proposed here is based on a single extractant, the 1,3,5-trimethoxy-2,4,6-tricarboxy-p-tert-butylcalix[6]arene, used to selectively separate U, Pu and Am from the urinary matrix prior to be measured. Using this analytical protocol, U and Pu are quantitatively and selectively recovered in two different acidic backextraction solutions whereas Am is quantitatively and selectively recovered in the organic phase. Furthermore, the purification stage is considerably shortened. The uranium and plutonium amounts are measured in aqueous phases using alpha spectrometry or inductively coupled plasma-mass spectrometry, whereas Am is measured in the organic phase using alpha liquid scintillation (photon/electron-rejecting alpha liquid scintillation).

The experimental procedure used for the indirect measurement of internal exposure of workers involved in the first step of the decontamination of a nuclear power plant which is being decommissioned is presented. The establishment of decontamination and decommissioning programs is currently in progress at the nuclear power plant and monitoring procedures to assess workers' exposure have also been implemented. Due to the presence of transuranics, the monitoring program includes the analysis of excreta samples. An analytical method for the determination of plutonium (Pu) and americum (Am) in urine samples was developed. The radiochemical separation is based on the coprecipitation of transuranics and sequential isolation of Pu and Am by anion exchange and extraction chromatography. Finally, electrodeposited sources are prepared and counted by alpha-spectrometry.

The primary goal of radiation protection in decommissioning and decontamination of the old nuclear facilities of the CIEMAT is to monitor and minimize exposure of personnel. Monitoring programs include determination of actinides and 90Sr in biological samples. A technique for the sequential measurement of low levels of 239Pu, 241Am and 90Sr in urine samples has been developed. The method involves coprecipitation of these radionuclides as phosphates from bulk urine sample. Separation of Plutonium is carried out using a conventional anion exchange technique. Americium and strontium isolations are achieved sequentially by chromatographic extraction (Tru.Spec and Sr.Spec columns) from the load and rinse solutions coming from the anion exchange column. Plutonium and Americium measurements are performed by alpha spectrometry. The mean recovery obtained is 80% and the detection limit for 24 h urine sample (1.41) is 0.6 mBq L-1. 90Sr determination is made by liquid scintillation counting. The detection limit in this case is 1.1 E-01 Bq/L.

A follow-up to our previous study is described on human gut transfer factors for plutonium and americium in winkles harvested from the shoreline near the British Nuclear Fuels plc Sellafield reprocessing plant. Six male and two female adult volunteers ate single quantities of shellfish; samples of their urine before and after consumption were analysed. As in the previous study, the plutonium and americium excreted were difficult to detect, and counting times of approximately 8 weeks were employed. Results are presented in detail to allow further interpretation, but preliminary assessment has been carried out on the basis of published data for excretion following transfer to blood. The mean and range obtained for the gut transfer for plutonium was 1.3 (0.2-3.2) x 10(-4); for americium the result was 0.8 (0.4-1.7) x 10(-4). Combined with the results of the previous study when interpreted on the same basis, the results were 1.7 (0.2-4.9) x 10(-4) and 0.9 (0.3-2.8) x 10(-4), respectively. No significant evidence was revealed in this study for differences in gut transfer between males and females, nor for seasonal differences, e.g. due to differences in form of incorporation of actinides in shellfish. For realistic assessments of doses due to shellfish consumption near Sellafield, to assess compliance with criteria recommended by the International Commission on Radiological Protection (ICRP), a suitable value for the gut transfer factor of both plutonium and americium might be 2 x 10(-4). A value of 5 x 10(-4) would appear to be cautious.

Intake of 241Am was reported in a young female technologist. External monitoring, whole body counting, urinalysis and fecal analysis were performed to determine the subsequent fate of the contaminant. Five days later, more than 99.5% of the radioactivity was voided in a fecal sample. A single particle, containing 2.85 microCi of 241AmO2 incorporated in a ceramic matrix, was isolated from the fecal sample. Brief descriptions of the radioanalytical results and dosimetry implications are presented. A shadow shield whole body counter was conveniently used to make an early estimate of the intake. This initial estimate enabled staff to decide that it was not necessary to artifically remove the contaminant. It was estimated that the lower large intestine was the organ which received the highest dose due to the passage of the ceramic particle. Systemic uptake of 241Am was indicated by urinalysis. The fractional transfer of 241Am from the GI tract to the blood was estimated to be less than 6 X 10(-5). This maximum estimate is limited by the MDA of the analytical procedures used. The fractional transfer of the 241Am from the GI tract to blood in this case is about an order of magnitude less than recent ICRP recommendations for "all compounds of americium".

A study was made of the role of biocomponents of bile, urine and feces in the elimination of plutonium and americium from the organism. Plutonium 239 and americium 241 were separated in bile due to higher tropism of plutonium to low molecular weight ligands, and of americium, to a protein-containing fraction. The status of plutonium excreted in feces was the same as the physicochemical status of americium. Plutonium 239 and americium 241 eliminated in urine were in a completely ultrafiltered state.

Data on gut transfer factors for environmental forms of radionuclides are essential for estimates of public radiation exposures following ingestion, and thus in decisions on controlling waste discharges. Dose estimates for transuranic nuclides are particularly sensitive to uncertainties stemming from gut transfer data being related to non-environmental forms and/or derived from animal experiments. We have measured human gut transfer factors for plutonium and americium in two experiments using marine foods obtained near Sellafield, Cumbria. Firstly, the urine of volunteer members of the critical group of shellfish consumers was analysed for transuranics and the results related to their consumption rates. Secondly, remotely-based volunteers ate single quantities of shellfish obtained near Sellafield, and their urine was analysed. An overall result for the gut transfer factor for environmental plutonium of 0.8 X 10(-4) indicates no need to increase the value of 1 X 10(-4), currently used by the International Commission on Radiological Protection (ICRP) for soluble forms. Results for americium show that the ICRP value of 5 X 10(-4) is maximising, and that a value of 1 X 10(-4) would be supportable. The results from the study of critical group members provide confidence in our habits survey techniques and reassurance that there are no significant pathways for intake of transuranics by these people that have not been recognised.

The estimated systemic burden of 0.3 muCi 241Am, made by the resident health physicist, has been compared to postmortem measurements using in-vivo counting equipment and radiochemical analysis of the donor's body. The health physicist's estimate was based on in-vivo measurements and urine excretion data obtained while the donor was still working, and it was logically assumed that the intake occurred about 1954 as a result of inhalation of Am-contaminated air. Based on additional data obtained by postmortem analysis, it seems more reasonable to assume for the purpose of scientific study and evaluation that the intake occurred in 1953, primarily as the result of a contaminated wound. Evaluations of the systemic burden using that assumption and three models indicate that the amount initially deposited could have been in the range of 0.2-1.1 muCi 241Am.

A technique has been developed for the determination of americium 241 in urine by a radiochemical purification of the nuclide from uranium (upon co-precipitation of americium 241 with calcium and lanthanum), plutonium, thorium, and polonium 210 (upon co-precipitation of these radionuclides with zirconium iodate). alpha-Radioactivity was measured either in a thick layer of the americium 241 precipitate with a nonisotope carrier or in thin-layer preparations after electrolytic precipitation of americium 241 on a cathode.