Urine Bioassay Research Articles

Evaluating the performance of the ORTEC(R) DetectiveTM for emergency urine bioassay

The performance of the ORTEC(®) Detective™ as a field deployable tool for emergency urine bioassay of (137)Cs, (60)Co, (192)Ir, (169)Yb and (75)Se was evaluated against ANSI N13.30. The tested activity levels represent 10 % RL (reference level) and 1 % RL defined by [Li C., Vlahovich S., Dai X., Richardson R. B., Daka J. N. and Kramer G. H. Requirements for radiation emergency urine bioassay techniques for the public and first responders. Health Phys (in press, 99(5), 702-707 (2010)]. The tests were conducted for both single radionuclide and mixed radionuclides at two geometries, one conventional geometry (CG) and one improved geometry (IG) which improved the MDAs (minimum detectable amounts) by a factor of 1.6-2.7. The most challenging radionuclide was (169)Yb. The measurement of the mixture radionuclides for (169)Yb at the CG did not satisfy the ANSI N13.30 requirements even at 10 % RL. At 1 % RL, (169)Yb and (192)Ir were not detectable at either geometry, while the measurement of (60)Co in the mixed radionuclides satisfied the ANSI N13.30 requirements only at the IG.

REQUIREMENTS FOR RADIATION EMERGENCY URINE BIOASSAY TECHNIQUES FOR THE PUBLIC AND FIRST RESPONDERS

Following a radiation emergency, the affected public and the first responders may need to be quickly assessed for internal contamination by the radionuclides involved. Urine bioassay is one of the most commonly used methods for assessing radionuclide intake and radiation dose. This paper attempts to derive the sensitivity requirements (from inhalation exposure) for the urine bioassay techniques for the top 10 high-risk radionuclides that might be used in a terrorist attack. The requirements are based on a proposed reference dose to adults of 0.1 Sv (CED, committed effective dose). In addition, requirements related to sample turnaround time and field deployability of the assay techniques are also discussed. A review of currently available assay techniques summarized in this paper reveals that method development for ²⁴¹Am, ²²⁶Ra, ²³⁸Pu, and ⁹⁰Sr urine bioassay is needed.

Is alpha spectrometry reliable for 210Po urine bioassay?

Typically the bioassay method for (210)Po in urine by alpha spectrometry (AS) involves wet decomposition of the sample, which may cause a loss of (210)Po if volatile species are present. To test this hypothesis, urine samples collected from two rats that were i.v. administered with polonium citrate were measured by both AS and liquid scintillation counting, where urine samples were mixed with a scintillation cocktail without any treatment. A split-plot design method was used to compare results from the two measurement methods, showing no evidence of a difference between the two methods. This suggests that the AS method is reliable for (210)Po urine bioassay.

A CASE OF WOUND INTAKE OF PLUTONIUM ISOTOPES AND 241Am IN A HUMAN: APPLICATION AND IMPROVEMENT OF THE NCRP WOUND MODEL

Plutonium isotopes (239Pu and 238Pu, and 241Am) with a total activity of 269 kBq were accidentally deposited in a puncture wound of the right index finger of a nuclear worker at the Mayak Production Association. Tissues surrounding the wound site contaminated with radionuclides were excised 4.5 h after the injury. Residual contamination within the wound amounted to 0.05% of the initial contamination. The 10-d therapy with CaNa3-diethylene triamine pentaacetate acid (CaNa3-DTPA) was performed in parallel with in vivo measurements of the wound site and daily urine bioassays. The wound intake of radionuclides was consistent with two forms of radioactive materials detected within the wound site, i.e., soluble compounds and a large fragment, which was completely removed by excision. On day 9 after the injury, the clearance rate from the wound site was 1.8 times higher than the rate predicted by the National Council on Radiation Protection and Measurements (NCRP) model for soluble compounds of plutonium and americium. The NCRP model parameters of transfer rates from the colloid and intermediate state (CIS) into soluble, and particles, aggregates and bound state (PABS) compartments were modified to eliminate any difference. As a result, a difference between the observed wound site radionuclide content and the value predicted by the modified wound model did not exceed 14% up to 9 days after the injury. For a longer period from 7 to 24 months, the value predicted by the modified model was consistent with results of the corresponding in vivo measurements. The treatment reduced the effective dose (50 years) from internal exposure by at least 480 times. The dose estimated (without accounting for a contribution of exposure dose to the regional lymph nodes draining the wound site) did not exceed 11 mSv.

Estimation of biological half-life of tritium in coastal region of India

The present study estimates biological half-life (BHL) of tritium by analysing routine bioassay samples of radiation workers. During 2007-2009 year, 72,100 urine bioassay samples of the workers were analysed by liquid scintillation counting technique for internal dose monitoring for tritium. Two hundred and two subjects were taken for study with minimum 3 μCiL(-1) tritium uptake in their body fluid. The BHL of tritium of subjects ranges from 1 to 16 d with an average of 8.19 d. Human data indicate that the biological retention time ranges from 4 to 18 d with an average of 10 d. The seasonal variations of the BHL of tritium are 3.09 ± 1.48, 6.87 ± 0.58 and 5.73 ± 0.76 d (mean ± SD) for summer, winter and rainy seasons, respectively, for free water tritium in the coastal region of Karnataka, India, which shows that the BHL in summer is twice that of the winter season. Also three subjects showed the BHL of 101.73-121.09 d, which reveals that organically bound tritium is present with low tritium uptake also. The BHL of tritium for all age group of workers is observed independent of age and is shorter during April to May. The distribution of cumulative probability vs. BHL of tritium shows lognormal distribution with a geometric mean of 9.11 d and geometric standard deviation of 1.77 d. The study of the subjects is fit for two-compartment model and also an average BHL of tritium is found similar to earlier studies.

Alimentary tract absorption (f1 values) for radionuclides in local and regional fallout from nuclear tests.

This paper presents gastrointestinal absorption fractions (f1 values) for estimating internal doses from local and regional fallout radionuclides due to nuclear tests. The choice of f1 values are based on specific circumstances of weapons test conditions and a review of reported f1 values for elements in different physical and chemical states. Special attention is given to fallout from nuclear tests conducted at the Marshall Islands. We make a distinction between the f1 values for intakes of radioactive materials immediately after deposition (acute intakes) and intakes that occur in the course of months and years after deposition, following incorporation into terrestrial and aquatic foodstuffs (chronic intakes). Multiple f1 values for different circumstances where persons are exposed to radioactive fallout (e.g., local vs. regional fallout and coral vs. continental tests) are presented when supportive information is available. In some cases, our selected f1 values are similar to those adopted by the International Commission on Radiological Protection (ICRP) (e.g., iodine and most actinides). However, f1 values for cesium and strontium derived from urine bioassay data of the Marshallese population are notably lower than the generic f1 values recommended by ICRP, particularly for acute intakes from local fallout (0.4 and 0.05 for Cs and Sr, respectively). The f1 values presented here form the first complete set of values relevant to realistic dose assessments for exposure to local or regional radioactive fallout.

Acute and chronic intakes of fallout radionuclides by Marshallese from nuclear weapons testing at Bikini and Enewetak and related internal radiation doses.

Annual internal radiation doses resulting from both acute and chronic intakes of all important dose-contributing radionuclides occurring in fallout from nuclear weapons testing at Bikini and Enewetak from 1946 through 1958 have been estimated for the residents living on all atolls and separate reef islands of the Marshall Islands. Internal radiation absorbed doses to the tissues most at risk to cancer induction (red bone marrow, thyroid, stomach, and colon) have been estimated for representative persons of all population communities for all birth years from 1929 through 1968, and for all years of exposure from 1948 through 1970. The acute intake estimates rely on a model using, as its basis, historical urine bioassay data, for members of the Rongelap Island and Ailinginae communities as well as for Rongerik residents. The model also utilizes fallout times of arrival and radionuclide deposition densities estimated for all tests and all atolls. Acute intakes of 63 radionuclides were estimated for the populations of the 20 inhabited atolls and for the communities that were relocated during the testing years for reasons of safety and decontamination. The model used for chronic intake estimates is based on reported whole-body, urine, and blood counting data for residents of Utrik and Rongelap. Dose conversion coefficients relating intake to organ absorbed dose were developed using internationally accepted models but specifically tailored for intakes of particulate fallout by consideration of literature-based evidence to choose the most appropriate alimentary tract absorption fraction (f1) values. Dose estimates were much higher for the thyroid gland than for red marrow, stomach wall, or colon. The highest thyroid doses to adults were about 7,600 mGy for the people exposed on Rongelap; thyroid doses to adults were much lower, by a factor of 100 or more, for the people exposed on the populated atolls of Kwajalein and Majuro. The estimates of radionuclide intake and internal radiation dose to the Marshallese that are presented in this paper are the most complete available anywhere and were used to make projections of lifetime cancer risks to the exposed populations, which are presented in a companion paper in this volume.

Method comparison for 241Am emergency urine bioassay

241Am is one of the high-risk radionuclides that might be used in a terrorist attack. 241Am in urine bioassay can identify the contaminated individuals who need immediate medical intervention and decontamination. This paper compares three methods for the measurement of 241Am in urine, namely liquid scintillation counting (LSC), inductively coupled plasma mass spectrometry (ICP-MS) and gamma spectrometry (GS), at two levels, 20 and 2 Bq l(-1). All three methods satisfied the ANSI N13.30 radio-bioassay criteria for accuracy and repeatability. ICP-MS offered the best sensitivity and fastest sample turnaround; however, the ICP-MS system used in this work may not be available in many bioassay laboratories. LSC and GS are more commonly available instruments. GS requires minimal or no sample preparation, which makes it a good candidate method. Moreover, the sample throughput can be significantly improved if the GS and LSC methods are automated.

An emergency urine bioassay method for 241Am by extraction chromatography and liquid scintillation counting

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.

Determination of an Environmental Background Level of 90Sr in Urine for the Hanford Bioassay Program

During the decommissioning and maintenance of some of the facilities at the U.S. Department of Energy Hanford Site in Washington State, workers have potential for a Sr intake. However, because of worldwide radioactive fallout, Sr is present in our environment and can be detectable in routine urine bioassay samples. It is important for the Hanford Site bioassay program to discriminate an occupational intake from a non-occupational environmental one. A detailed study of the background Sr in the urine of unexposed Hanford workers was performed. A survey of the Hanford Site bioassay database found 128 Hanford workers who were hired between 1997 and 2002 and who had a very low potential for an occupational exposure prior to the baseline strontium urinalysis. Each urinalysis sample represented excretion during an approximate 24-h period. The arithmetic mean value for the 128 pre-exposure baselines was 3.6 +/- 5.1 mBq d. The 99 percentile result was 17 mBq d, which was interpreted to mean that 1% of Hanford workers not occupationally exposed to strontium might exceed 17 mBq d.

Uncertainties of Mayak urine data

A method of parameterising the likelihood functions representing the uncertainty of Mayak plutonium urine bioassay measurements is described. The Poisson-lognormal model is assumed and data from 63 cases (1,087 urine measurements in all) are used to empirically determine the lognormal normalisation uncertainty, given the measurement uncertainties obtained from count quantities. An outlier-insensitive procedure is used to fit the cumulative probability distribution of scaled deviations in order to determine the normalisation uncertainty. The natural logarithm of the geometric standard deviation of the total normalisation uncertainty is found to be 0.34 including a measurement component estimated to be 0.2.

Noninvasive quantitation of human liver steatosis using magnetic resonance and bioassay methods

The purpose was to evaluate the ability of three magnetic resonance (MR) techniques to detect liver steatosis and to determine which noninvasive technique (MR, bioassays) or combination of techniques is optimal for the quantification of hepatic fat using histopathology as a reference. Twenty patients with histopathologically proven steatosis and 24 control subjects underwent single-voxel proton MR spectroscopy (MRS; 3 voxels), dual-echo in phase/out of phase MR imaging (DEI) and diffusion-weighted MR imaging (DWI) examinations of the liver. Blood or urine bioassays were also performed for steatosis patients. Both MRS and DEI data allowed to detect steatosis with a high sensitivity (0.95 for MRS; 1 for DEI) and specificity (1 for MRS; 0.875 for DEI) but not DWI. Strong correlations were found between fat fraction (FF) measured by MRS, DEI and histopathology segmentation as well as with low density lipoprotein (LDL) and cholesterol concentrations. A Bland-Altman analysis showed a good agreement between the FF measured by MRS and DEI. Partial correlation analyses failed to improve the correlation with segmentation FF when MRS or DEI data were combined with bioassay results. Therefore, FF from MRS or DEI appear to be the best parameters to both detect steatosis and accurately quantify fat liver noninvasively.

Uncertainties in internal doses calculated for mayak workers--a study of 63 cases

This study makes use of 63 cases of Mayak workers exposed to Pu-239 with autopsy data and some late-time urine bioassay data. In addition, air-concentration data--used to construct monthly average values--are available for each case, which provide the time dependence and potential magnitudes of normal inhalation intakes for each case. The purpose of the study is to develop and test Bayesian methods of dose calculation for the Mayak workers. The first part of the study was to quantitatively characterise the uncertainties of the bioassay data. Then, starting with three different published biokinetic models, the data are fit by varying intake and model perturbation parameters, e.g., parameters influencing the lung, thoracic lymph nodes, liver and bone retention. Statistical self-consistency arguments are used to check the measurement uncertainty parameters within the Poisson-lognormal model. The second part of the study is to set up and test Bayesian dose calculations, which use the point determinations of biokinetic parameters from the study cases within a discrete, empirical Bayes approximation. The main conclusion of the study is that these methods are now ready to be applied to the entire Mayak worker population.

A study of early Los Alamos internal exposures to plutonium

Internal dose caused by exposure to (239)Pu/(240)Pu is calculated for a group of 210 former Los Alamos workers who participated in the urine bioassay programme during the years 1944-45. An iterative Bayesian procedure is employed, where the distribution of intake amounts resulting from an initial calculation is used to define a prior probability distribution of inhalation intakes for an iterated second calculation. The urine bioassay data from this time period were not of high quality, and the more accurate intake prior tempers the effect of spurious high samples, which were probably caused by sample contamination.

Method intercomparison for the analysis of 239/240Pu in human urine

Following a radiological or nuclear emergency, medical intervention requires rapid assessment of the exposure of people usually through determination of internal dose. For the plutonium urine bioassay, besides thermal ionization mass spectrometry (TIMS) and alpha spectrometry methods, inductively coupled plasma mass spectrometry (ICP-MS) methods have been recently developed, which can provide much higher sample throughput. In this work, three ICP-MS methods were compared with one TIMS method and two alpha spectrometry methods for the measurement of 239Pu and 240Pu in human urine samples spiked at different concentration levels. The sample throughputs for all three ICP-MS methods are similar: each instrument measures about 80 samples in the first 24 hours and 200 samples in the first 48 hours following the emergency event, if the samples arrive at the laboratory 8 hours after the event occurs. Method accuracy and precision were determined using ANSI N13.30. Method detection limits and minimum detectable amounts (MDA) were determined to evaluate method sensitivities. The sensitivities of the three ICP-MS methods were also compared with the derived urine action level (24 h urine, 500 mSv committed effective dose equivalent, inhalation exposure, maximum dose conversion factor) to evaluate their applicability to exposure situations.

UNCERTAINTIES ANALYSIS FOR THE PLUTONIUM DOSIMETRY MODEL, DOSES-2005, USING MAYAK BIOASSAY DATA

The Doses-2005 model is a combination of the International Commission on Radiological Protection (ICRP) models modified using data from the Mayak Production Association cohort. Surrogate doses from inhaled plutonium can be assigned to approximately 29% of the Mayak workers using their urine bioassay measurements and other history records. The purpose of this study was to quantify and qualify the uncertainties in the estimates for radiation doses calculated with the Doses-2005 model by using Monte Carlo methods and perturbation theory. The average uncertainty in the yearly dose estimates for most organs was approximately 100% regardless of the transportability classification. The relative source of the uncertainties comes from three main sources: 45% from the urine bioassay measurements, 29% from the Doses-2005 model parameters, and 26% from the reference masses for the organs. The most significant reduction in the overall dose uncertainties would result from improved methods in bioassay measurement with additional improvements generated through further model refinement. Additional uncertainties were determined for dose estimates resulting from changes in the transportability classification and the smoking toggle. A comparison was performed to determine the effect of using the model with data from either urine bioassay or autopsy data; no direct correlation could be established. Analysis of the model using autopsy data and incorporation of results from other research efforts that have utilized plutonium ICRP models could improve the Doses-2005 model and reduce the overall uncertainty in the dose estimates.

Investigation of monitoring for internal exposure by urine bioassay in a biomedical research facility

A simple monitoring programme by urine bioassay was carried out to assess internal doses for workers in a biomedical research facility. Urine samples were measured without chemical treatment using a liquid scintillation counter or sodium iodide (NaI (Tl)) scintillation detector. The detection limits for the committed effective doses were below 1 mSv for 125I and 131I and below 1 x 10(-1) mSv for 3H, 14C, 32P, 35S and 51Cr, in the case that samples were collected within 3 d after handling. All of the urinary concentrations were below the detection limit, suggesting that no significant intake was detected during the investigation. The present results suggest that personal monitoring, such as the urine bioassay, is not necessary in many cases as a routine monitoring for workers engaged in tracer experiments using the above nuclides.

Usefulness of the simultaneous bioassay analysis method used for an intake estimation of a radionuclide

An intake of a radionuclide is estimated based on bioassay measurement data obtained by an in vivo or an in vitro method. Often the intake estimates from one bioassay analysis are considerably different from other results. For better estimates, a simultaneous or combined analysis of measurement data from different bioassay methods is attempted. In this study, the usefulness of a simultaneous bioassay analysis was investigated by using the IDEAS/IAEA intercomparison exercise data and the Individual Monitoring of an Internal Exposure computer code. Tests were made for whole-body counting and urine assay against an acute inhalation of types M and S 60Co particles with various activity median aerodynamic diameter (AMAD). The data set excluding rogue data as well as all the available data were used in this study. The best estimated intake was evaluated based on the best-fit time of an intake determined by minimizing the mean relative deviation Dr. In the case of the whole-body and urine bioassay by using the data excluding some rogue data, the smallest Dr appears at 0.1 and 10 microm of AMAD, respectively, which are different from those estimated by using all the available data. In the case of the simultaneous analysis, it appears at 20 microm of AMAD, which is the same as that estimated by using all the available data. Supposing that monitoring data of a good quality is available, it is expected that the application of a simultaneous analysis to different bioassay methods can provide not only better estimates of an intake but also insights into the validity of the models and parameters used in an interpretation.

Comparison of dose estimation from occupational exposure to 239Pu using different modelling approaches

Several approaches are available for bioassay interpretation when assigning Pu doses to Mayak workers. First, a conventional approach is to apply ICRP models per se. An alternative method involves individualised fitting of bioassay data using Bayesian statistical methods. A third approach is to develop an independent dosimetry system for Mayak workers by adapting ICRP models using a dataset of available bioassay measurements for this population. Thus, a dataset of 42 former Mayak workers, who died of non-radiation effects, with both urine bioassay and post-mortem tissue data was used to test these three approaches. All three approaches proved to be adequate for bioassay and tissue interpretation, and thus for Pu dose reconstruction purposes. However, large discrepancies are observed in the resulting quantitative dose estimates. These discrepancies can, in large part, be explained by differences in the interpretation of Pu behaviour in the lungs in the context of ICRP lung model. Thus, a careful validation of Pu lung dosimetry model is needed in Mayak worker dosimetry systems.

Discussion of 241am role as the factor of internal radiation exposure in Mayak PA workers

Radiation monitoring with whole body counter (WBC) has a series of advantages, as this technique is efficient and is considerably cheaper than indirect urine bioassay method. Experience of WBC operation in the system of Mayak PA worker internal radiation monitoring as well as qualitative analysis of the obtained results is considered. The groups of workers under study are described, and a brief analysis of data obtained in comparison with individual exposure parameters is presented. The number of individuals with 241Am is higher among the workers employed in the early years of Mayak PA operation. Fraction of 241Am in total body burden relative to sum of actinides is higher for workers who started to work in conditions of exposure in the last decade (1990-2000) and sometimes is as high as 25%.