Physical aspects of scintigraphybased dosimetry for nuclear medicine therapy

In nuclear medicine therapy the treatment of tumours by radiation exposure from internally deposited labelled antibodies or labelled peptides is currently an active field of investigation. To permit the efficient delivery of high amounts of radiation dose to tumours while limiting the radiation dose to critical organs dosimetry calculations have to be performed. These are relying on scintigraphic data being input to the well known MIRD formalism.

This paper focuses on the methods and the difficulties associated with the scintigraphic determination of organ kinetics. The physical properties of the well-known scintigraphic imaging modalities, PET, SPECT and planar scintigraphy, are discussed thereby taking into account the properties of the appropriate radionuclides currently being available for therapy and dosimetry. Several arguments are given and disputed for the limited clinical use of PET and SPECT in dosimetry and the ongoing preference of planar whole-body imaging as the method of choice. The quantitative restrictions still inherent to this method are also discussed in detail. Procedural recommendations are proposed covering all processes related to data acquisition, data correction and data analysis which finally lead to reliable estimations of organ dose.

Evaluation of in vivo and in vitro dose detection limits for different radionuclides and measurement techniques

Personal monitoring programs for workers handling radioactive materials are influenced by numerous factors as the measurements of radioactivity in tissues or/and in excreta can be carried out using different techniques. This paper summaries the basic procedures needed for accurate and fast measurement of different radionuclides like (235)U, (234)U, (238)U, (226)Ra, (210)Po, (131)I, (99m)Tc, (134)Cs, (137)Cs, (57)Co, (58)Co, and (60)Co. Overviews of in vitro and in vivo monitoring methods are provided as well as methods used to calculate detection limits and internal radiation dose. For the radionuclides of interest, in vivo and in vitro detection limits were converted into committed effective doses to evaluate the applicability and limitations of the systems used at the laboratory. The results proved that the systems' sensitivity is suitable for use in routine monitoring of workers subject to risk of internal exposure from such radionuclides. Consequently, monitoring programs suggested by the Syrian internal dosimetry laboratory are suitable to detect committed effective doses even below 1mSv in most cases.

Excreta Sampling as an Alternative to In Vivo Measurements at the Hanford Site

The capabilities of indirect radiobioassay by urine and fecal sample analysis were compared with the direct radiobioassay methods of whole body counting and lung counting for the most common radionuclides and inhalation exposure scenarios encountered by Hanford workers. Radionuclides addressed by in vivo measurement included 137Cs, 60Co, 154Eu, and 241Am as an indicator for plutonium mixtures. The same radionuclides were addressed using gamma energy analysis of urine samples, augmented by radiochemistry and alpha spectrometry methods for plutonium in urine and fecal samples. It was concluded that in vivo whole body counting and lung counting capability should be maintained at the Hanford Site for the foreseeable future, however, urine and fecal sample analysis could provide adequate, though degraded, monitoring capability for workers as a short-term alternative, should in vivo capability be lost due to planned or unplanned circumstances.

Determination of 234U/238U, 235U/238U and 236U/238U isotope ratios in urine using sector field inductively coupled plasma mass spectrometry

Quantification of the isotopic composition of uranium in urine at low levels of concentration is important for assessing both military and civilian populations' exposures to uranium. However, until now there has been no convenient, precise method established for rapid determination of multiple uranium isotope ratios. Here, the authors report a new method to measure (234)U/(238)U, (235)U/(238)U and (236)U/(238)U. It uses solid-phase chelation extraction (via TRU columns) of actinides from the urine matrix, followed by measurement using a magnetic sector field inductively coupled plasma mass spectrometer (SF-ICP-MS-Thermo Element XR) equipped with a high-efficiency nebulizer (Apex PFA microflow) and coupled with a membrane desolvating nebulizer system (Aridus II™). This method provides rapid and reliable results and has been used successfully to analyse Certified Reference Materials.

Toxicity of irradiated advanced heavy water reactor fuels

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

Reliability of a new biokinetic model of zirconium in internal dosimetry: part I, parameter uncertainty analysis

The reliability of biokinetic models is essential in internal dose assessments and radiation risk analysis for the public, occupational workers, and patients exposed to radionuclides. In this paper, a method for assessing the reliability of biokinetic models by means of uncertainty and sensitivity analysis was developed. The paper is divided into two parts. In the first part of the study published here, the uncertainty sources of the model parameters for zirconium (Zr), developed by the International Commission on Radiological Protection (ICRP), were identified and analyzed. Furthermore, the uncertainty of the biokinetic experimental measurement performed at the Helmholtz Zentrum München-German Research Center for Environmental Health (HMGU) for developing a new biokinetic model of Zr was analyzed according to the Guide to the Expression of Uncertainty in Measurement, published by the International Organization for Standardization. The confidence interval and distribution of model parameters of the ICRP and HMGU Zr biokinetic models were evaluated. As a result of computer biokinetic modelings, the mean, standard uncertainty, and confidence interval of model prediction calculated based on the model parameter uncertainty were presented and compared to the plasma clearance and urinary excretion measured after intravenous administration. It was shown that for the most important compartment, the plasma, the uncertainty evaluated for the HMGU model was much smaller than that for the ICRP model; that phenomenon was observed for other organs and tissues as well. The uncertainty of the integral of the radioactivity of Zr up to 50 y calculated by the HMGU model after ingestion by adult members of the public was shown to be smaller by a factor of two than that of the ICRP model. It was also shown that the distribution type of the model parameter strongly influences the model prediction, and the correlation of the model input parameters affects the model prediction to a certain extent depending on the strength of the correlation. In the case of model prediction, the qualitative comparison of the model predictions with the measured plasma and urinary data showed the HMGU model to be more reliable than the ICRP model; quantitatively, the uncertainty model prediction by the HMGU systemic biokinetic model is smaller than that of the ICRP model. The uncertainty information on the model parameters analyzed in this study was used in the second part of the paper regarding a sensitivity analysis of the Zr biokinetic models.

Evaluating the performance of the ORTEC® Detective™ 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.

Investigation of magnetic nanoparticles for the rapid extraction and assay of alpha-emitting radionuclides from urine: demonstration of a novel radiobioassay method

In the event of an accidental or intentional release of radionuclides into a populated area, massive numbers of people may require radiobioassay screening as triage for dose-reduction therapy or identification for longer-term follow-up. If the event released significant levels of beta- or alpha-emitting radionuclides, in vivo assays would be ineffective. Therefore, highly efficient and rapid analytical methods for radionuclide detection from submitted spot urine samples (≤50 mL) would be required. At present, the quantitative determination of alpha-emitting radionuclides from urine samples is highly labor intensive and requires significant time to prepare and analyze samples. Sorbent materials that provide effective collection and enable rapid assay could significantly streamline the radioanalytical process. The authors have demonstrated the use of magnetic nanoparticles as a novel method of extracting media for four alpha-emitting radionuclides of concern (polonium, radium, uranium and americium) from chemically-unmodified and pH-2 human urine. Herein, the initial experimental sorption results are presented along with a novel method that uses magnetic nanoparticles to extract radionuclides from unmodified human urine and then collect the magnetic field-induced particles for subsequent alpha-counting-source preparation. Additionally, a versatile human dose model is constructed that determines the detector count times required to estimate dose at specific protective-action thresholds. The model provides a means to assess a method's detection capabilities and uses fundamental health physics parameters and actual experimental data as core variables. The modeling shows that, with effective sorbent materials, rapid screening for alpha-emitters is possible with a 50-mL urine sample collected within 1 wk of exposure/intake.

Analysis of urine for pure beta emitters: methods and application

Bioassay for individual radionuclides is an essential and first step in estimation of radiation risk to nuclear facilities workers and people who are exposed to the contaminated environment in the event of a nuclear accident or radiological attack. Urine is a frequently used biological sample for this purpose. Tritium and (14)C are important radionuclides for workers in nuclear reactors and radiopharmaceutical laboratories. A method for the determination of tritium and (14)C in organic and inorganic forms in urine has been developed. It involves activated charcoal absorption of organic matter followed by combustion to separate tritiated water from organically-bound tritium. Inorganic (14)C from organically-bound (14)C, the separated tritium and (14)C were measured using liquid scintillation counting. Iodine-129, a long-lived beta emitter, is normally released to the atmosphere during the operation of nuclear facilities, especially in reprocessing plants. The high concentration of iodine in the thyroid makes this radionuclide an important source of exposure to exposed populations. A simple method has been developed in this work for the determination of (129)I in urine by anion exchange preconcentration, extraction purification and liquid scintillation counting. Using accelerator mass spectrometry, urine samples can be analyzed for low level (129)I in both organic and inorganic forms after active charcoal adsorption and solvent extraction separation. Condensed water collected daily from the reactor hall in a Danish research reactor and monthly urine samples from the staff working in the reactor building were collected from 2003-2010 and analyzed using this method, and the results are presented and discussed.

Rapid urinary output normalization method using specific gravity

Spot urine samples are often taken for emergency radiobioassay to provide a quick dose assessment for contaminated individuals. The subsequent dosimetric analysis requires a urinary output normalization method to adjust spot sampling to daily urine excretion. A rapid specific gravity method that was developed for 24-h urinary output correction is described. Spot urine samples were collected from volunteers of different race, gender, and age to validate the normalization method. Results show that a specific gravity test is a fast, easy and robust method to correct the urine excretion in the event of a radiation emergency.

Laboratory evaluation of a SpectraMax microplate reader and test strips for field measurement of creatinine in spot urine samples in the event of a radiological accident

The fear that terrorists might use radiological or nuclear (RN) devices to attack others is a new but growing phenomenon, arising mainly from the events of 11 September 2001. Research on rapid analytical methods that can allow analyses of large numbers of people who may become internally contaminated with radionuclides due to a RN accident is still limited. To contribute to this bioassay capacity for emergency response, the Radiation Protection Bureau of Health Canada has identified and evaluated two new portable SpectraMax plate readers (model 250 and Plus 384) and one brand of dry reagent strips for rapid measurement of creatinine in spot urine samples. Concentrations of creatinine in spot urine samples provide a means of adjusting or normalizing urine collections to 24 h, upon which accurate internal dose assessments due to the radionuclides can be made. Preliminary test results of the devices showed the two SpectraMax plate readers and the TECO dry creatinine reagent strips were portable, rapid and reliable for urinary creatinine measurements in spot samples, suggesting they can be used in rapid dose screening of people.

PROCORAD's proficiency tests as a means of analyzing capacities in emergency situations

Beyond the metrological evaluation indispensable to prove objectively the capability of a medical laboratory to perform analysis, the comparison of analytical performances regarding the specificity, rapidity and trueness of the available methods is a real interest of international comparison exercises. The objective of this paper is to present and discuss the main results of proficiency testing organized by PROCORAD in the radiobioassay field illustrated by exercises for in vitro "surprise" analysis when no information is available on the nuclides involved, which can be the case in an emergency situation.

Rapid methods for the isolation of actinides Sr, Tc and Po from raw urine

Rapid methods for the isolation and analysis of individual actinides (Th, U, Pu, Am/Cm) and Sr, Tc and Po from small volumes of raw urine have been developed. The methods involve acidification of the sample and the addition of aluminum nitrate or aluminum chloride salting-out agent prior to isolation of the desired analyte using a tandem combination of prefilter material and extraction chromatographic resin. The method has been applied to the separation of individual analytes from spiked urine samples. Analytes were recovered in high yield and radionuclide purity with separation times as low as 30 min. The chemistry employed is compatible with automation on the ARSIIe instrument.

A mobile bioassay laboratory for the assessment of internal doses based on in vivo and in vitro measurements

Internal exposures may occur in nuclear power plants, radioisotope production, and in medicine and research laboratories. Such practices require quick response in case of accidents of a wide range of magnitudes. This work presents the design and calibration of a mobile laboratory for the assessment of accidents involving workers and the population as well as for routine monitoring. The system was set up in a truck with internal dimensions of 3.30 m × 1.60 m × 1.70 m and can identify photon emitters in the energy range of 100-3,000 keV in the whole body, organs, and in urine. A thyroid monitor consisting of a lead-collimated NaI(Tl)3" × 3" (7.62 × 7.62 cm) detector was calibrated with a neck-thyroid phantom developed at the IRD (Instituto de Radioproteção e Dosimetria). Whole body measurements were performed with a NaI(Tl)8" × 4" (20.32 × 10.16 cm) detector calibrated with a plastic-bottle phantom. Urine samples were measured with another NaI(Tl) 3" × 3" (7.62 × 7.62 cm) detector set up in a steel support. Standard solutions were provided by the National Laboratory for Metrology of Ionizing Radiation of the IRD. Urine measurements are based on a calibration of efficiency vs. energy for standard volumes. Detection limits were converted to minimum committed effective doses for the radionuclides of interest using standard biokinetic and dosimetric models in order to evaluate the applicability and limitations of the system. Sensitivities for high-energy activation and fission products show that the system is suitable for use in emergency and routine monitoring of individuals under risk of internal exposure by such radionuclides.

Urinary excretion of radionuclides from Marshallese exposed to fallout from the 1954 Bravo nuclear test

Soon after the Bravo nuclear test at Bikini Atoll in the Marshall Islands on 1 March 1954, urine samples were collected for analysis of excreted radioactivity from native residents exposed to radioactive fallout on two atolls as well as from U.S. military personnel on a third atoll. The earliest acquired samples, obtained by the Los Alamos Scientific Laboratory (LASL), were assayed for various radionuclides and provided the first known measurements of (131)I in urine following exposure to fallout from a nuclear test. Over the course of 1954, many additional samples were collected by the LASL, as well as by the Atomic Energy Commission New York Operations Office's Health and Safety Laboratory and the Naval Radiological Defense Laboratory. Collectively, the groups sampled included Marshallese exposed on Rongelap and Ailinginae Atolls, American military weather observers temporarily resident on Rongerik Atoll, and sailors from the Japanese fishing vessel, the Lucky Dragon. While the bioassay measurement data and individual urine volumes have been crucial to various attempts to assess intakes of radioactivity and the related internal radiation doses among the Marshallese, those data have never been published in any peer-reviewed journal, but have been restricted to agency memoranda, laboratory reports, and summaries in some publications and book chapters. Reconstructions of internal doses to Marshallese in 1954 and in later years have depended on these data and, hence, they have considerable historical importance as well as importance to ongoing health risk projections for Marshallese. This paper presents much of the original data on urine volumes and radioactivity from the various assays of urine for radionuclides, and compares estimates of (131)I intakes made in 1954, 1985, 1987, and 2008.

What is the Objective of the Mass Balance Study? A Retrospective Analysis of Data in Animal and Human Excretion Studies Employing Radiolabeled Drugs

Mass balance excretion studies in laboratory animals and humans using radiolabeled compounds represent a standard part of the development process for new drugs. From these studies, the total fate of drug-related material is obtained: mass balance, routes of excretion, and, with additional analyses, metabolic pathways. However, rarely does the mass balance in radiolabeled excretion studies truly achieve 100% recovery. Many definitions of cutoff criteria for mass balance that identify acceptable versus unacceptable recovery have been presented as ad hoc statements without a strong rationale. To address this, a retrospective analysis was undertaken to explore the overall performance of mass balance studies in both laboratory animal species and humans using data for 27 proprietary compounds within Pfizer and extensive review of published studies. The review has examined variation in recovery and the question of whether low recovery was a cause for concern in terms of drug safety. Overall, mean recovery was greater in rats and dogs than in humans. When the circulating half-life of total radioactivity is greater than 50 h, the recovery tends to be lower. Excretion data from the literature were queried as to whether drugs linked with toxicities associated with sequestration in tissues or covalent binding exhibit low mass balance. This was not the case, unless the sequestration led to a long elimination half-life of drug-related material. In the vast majority of cases, sequestration or concentration of drug-related material in an organ or tissue was without deleterious effect and, in some cases, was related to the pharmacological mechanism of action. Overall, from these data, recovery of radiolabel would normally be equal to or greater than 90%, 85%, and 80% in rat, dog, and human, respectively. Since several technical limitations can underlie a lack of mass balance and since mass balance data are not sensitive indicators of the potential for toxicity arising via tissue sequestration, absolute recovery in humans should not be used as a major decision criteria as to whether a radiolabeled study has met its objectives. Instead, the study should be seen as an integral part of drug development answering four principal questions: 1) Is the proposed clearance mechanism sufficiently supported by the identities of the drug-related materials in excreta, so as to provide a complete understanding of clearance and potential contributors to interpatient variability and drug-drug interactions? 2) What are the drug-related entities present in circulation that are the active principals contributing to primary and secondary pharmacology? 3) Are there findings (low extraction recovery of radiolabel from plasma, metabolite structures indicative of chemically reactive intermediates) that suggest potential safety issues requiring further risk assessment? 4) Do questions 2 and 3 have appropriate preclinical support in terms of pharmacology, safety pharmacology, and toxicology? Only if one or more of these four questions remain unanswered should additional mass balance studies be considered.

Internal dose assessment of 210Po using biokinetic modeling and urinary excretion measurement

The mysterious death of Mr. Alexander Litvinenko who was most possibly poisoned by Polonium-210 ((210)Po) in November 2006 in London attracted the attention of the public to the kinetics, dosimetry and the risk of this high radiotoxic isotope in the human body. In the present paper, the urinary excretion of seven persons who were possibly exposed to traces of (210)Po was monitored. The values measured in the GSF Radioanalytical Laboratory are in the range of natural background concentration. To assess the effective dose received by those persons, the time-dependence of the organ equivalent dose and the effective dose after acute ingestion and inhalation of (210)Po were calculated using the biokinetic model for polonium (Po) recommended by the International Commission on Radiological Protection (ICRP) and the one recently published by Leggett and Eckerman (L&E). The daily urinary excretion to effective dose conversion factors for ingestion and inhalation were evaluated based on the ICRP and L&E models for members of the public. The ingestion (inhalation) effective dose per unit intake integrated over one day is 1.7 x 10(-8) (1.4 x 10(-7)) Sv Bq(-1), 2.0 x 10(-7) (9.6 x 10(-7)) Sv Bq(-1) over 10 days, 5.2 x 10(-7) (2.0 x 10(-6)) Sv Bq(-1) over 30 days and 1.0 x 10(-6) (3.0 x 10(-6)) Sv Bq(-1) over 100 days. The daily urinary excretions after acute ingestion (inhalation) of 1 Bq of (210)Po are 1.1 x 10(-3) (1.0 x 10(-4)) on day 1, 2.0 x 10(-3) (1.9 x 10(-4)) on day 10, 1.3 x 10(-3) (1.7 x 10(-4)) on day 30 and 3.6 x 10(-4) (8.3 x 10(-5)) Bq d(-1) on day 100, respectively. The resulting committed effective doses range from 2.1 x 10(-3) to 1.7 x 10(-2) mSv by an assumption of ingestion and from 5.5 x 10(-2) to 4.5 x 10(-1) mSv by inhalation. For the case of Mr. Litvinenko, the mean organ absorbed dose as a function of time was calculated using both the above stated models. The red bone marrow, the kidneys and the liver were considered as the critical organs. Assuming a value of lethal absorbed dose of 5 Gy to the bone marrow, 6 Gy to the kidneys and 8 Gy to the liver, the amount of (210)Po which Mr. Litvinenko might have ingested is therefore estimated to range from 27 to 1,408 MBq, i.e 0.2-8.5 microg, depending on the modality of intake and on different assumptions about blood absorption.

[Pharmacokinetic parameter and residua of 63Ni-NiCl2 in rat]

Absorption distribution and excretion of 63Ni-NiCl2, administered orally to rats were studied by using liquid scintillation counting method. It was observed that the concentration-time curves in blood fitted the two compartment model of pharmacokinetics, Ka=6.18 h(-1), T(1/2)alpha =0.79 h, T(1/2)beta=40.68 h, CL =0.42 mL kg(-1) h(-1), Tmax =0.53 h, Cmax=24,987.75 min(-1) mL(-1), and Vd=0.016 L kg(-1). After rats were treated by 63Ni-NiCl2 for 15 days, in 22 tissues tested, the contents of 63Ni-NiCl2 in hair, hypothalamus, hypophysis, pancreas, small and large intestines were higher, and the residua of 63Ni-NiCl2 was not discovered in liver, kidney and heart. Radioactivity eliminated was 83.27% by urine and feces, 54.86% by urine, 28.41% by feces.

Studies of a urinary biomarker of dietary inorganic sulphur in subjects on diets containing 1-38 mmol sulphur/day and of the half-life of ingested 34SO4(2-)

OBJECTIVE

Sulphites are widely used food additives that may damage health, hence limits are set on their use. They are excreted in urine as sulphate, along with sulphate derived from sulphur amino acids. Dietary intakes of sulphites are hard to determine, so we have tested the utility of urinary nitrogen:sulphate ratio as a biomarker of inorganic sulphur (IS) intake. Additionally we determined the half-life of ingested (34)SO(4)(2-) from its urinary excretion.

SUBJECTS

Twenty healthy adult subjects were recruited by poster advertisement, for a 24-h study where they ate specified foods, which were high in IS, in addition to their normal diet. The half-life of ingested (34)SO(4)(2-) was assessed in five healthy volunteers, given 5.9 mmols of Na(2)(34)SO(4) as a single dose and collecting all urine specimens for 72-96 h. Urine and duplicate diets from three previously conducted studies were analysed for nitrogen and sulphate content, thus expanding the range of IS intakes for evaluation.

METHODS

Duplicate diets were analysed for IS content by ion exchange chromatography, while IS intake was predicted from urinary sulphate (g/day S)-(urinary nitrogen (g/day)/18.89). (32)S:(34)S ratios were determined by liquid chromatography mass spectrometry/mass spectrometry.

RESULTS

The range of IS intake was 1.3-37.5 mmol S/day. Actual and predicted IS intakes were mmol/day+/-s.e. 9.2+/-0.65 and 7.0+/-0.45, respectively, and were correlated r=0.60 (n=108). The mean half-life of ingested (34)SO(4)(2-) was 8.2 h.

CONCLUSIONS

From a 24-h urine collection, IS intake from the habitual diet can be determined for groups of individuals. To predict individual intakes of IS, which may include high sporadic amounts from beer and wine, at least 48 h of urine collection would be required.

Effect of sodium [36Cl]chlorate dose on total radioactive residues and residues of parent chlorate in beef cattle

The objectives of this study were to determine total radioactive residues and chlorate residues in edible tissues of cattle administered at three levels of sodium [36Cl]chlorate over a 24-h period and slaughtered after a 24-h withdrawal period. Three sets of cattle, each consisting of a heifer and a steer, were intraruminally dosed with a total of 21, 42, or 63 mg of sodium [36Cl]chlorate/kg of body weight. To simulate a 24-h exposure, equal aliquots of the respective doses were administered to each animal at 0, 8, 16, and 24 h. Urine and feces were collected in 12-h increments for the duration of the 48-h study. At 24 h after the last chlorate exposure, cattle were slaughtered and edible tissues were collected. Urine and tissue samples were analyzed for total radioactive residues and for metabolites. Elimination of radioactivity in urine and feces equaled 20, 33, and 48% of the total dose for the low, medium, and high doses, respectively. Chlorate and chloride were the only radioactive chlorine species present in urine; the fraction of chlorate present as a percentage of the total urine radioactivity decreased with time regardless of the dose. Chloride was the major radioactive residue present in edible tissues, comprising over 98% of the tissue radioactivity for all animals. Chlorate concentrations in edible tissues ranged from nondetectable to an average of 0.41 ppm in skeletal muscle of the high-dosed animals. No evidence for the presence of chlorite was observed in any tissue. Results of this study suggest that further development of chlorate as a preharvest food safety tool merits consideration.

Tissue distribution, elimination, and metabolism of dietary sodium [36Cl]chlorate in beef cattle

Two steers (approximately 195 kg) were each dosed with 62.5 or 130.6 mg/kg body weight sodium [36Cl]chlorate for three consecutive days. All excreta were collected during the dosing and 8 h withdrawal periods. The apparent radiochlorine absorption was 62-68% of the total dose with the major excretory route being urine. Parent chlorate was 65-100% of the urinary radiochlorine; chloride was the only other radiochlorine species present. Similarly, residues in edible tissues were composed of chloride and chlorate with chloride being the major radiolabeled species present. Chlorate represented 28-57% of the total radioactive residues in skeletal muscle; in liver, kidney, and adipose tissues, chlorate ion represented a smaller percentage of the total residues. Chlorate residues in the low dose steer were 26 ppm in kidney, 14 ppm in skeletal muscle, 2.0 ppm in adipose tissue, and 0.7 ppm in liver. These data indicate that sodium chlorate may be a viable preharvest food safety tool for use by the cattle industry.

Biodistribution and excretion of radioactivity after the administration of 166Ho-chitosan complex (DW-166HC) into the prostate of rat

PURPOSE

The objective of this study was to determine the fate of the 166Ho-chitosan complex (DW-166HC) in rats by examining its absorption, distribution and excretion after administration into the prostate.

METHODS

About 100 microCi of DW-166HC [containing 0.1875 mg of Ho(NO3)3.5H2O and 0.25 mg of chitosan] was administered intraprostatically. The level of radioactivity in blood, urinary and faecal excretion, and radioactivity distribution were examined. To determine the effect of chitosan in DW-166HC, 166Ho nitrate alone [0.1875 mg of Ho(NO3)3.5H2O] was administered into the prostate of male rats, and radioactivity distribution was examined using whole-body autoradiography.

RESULTS

After administration of DW-166HC into the prostate, cumulative urinary and faecal excretion over the period 0-72 h was 0.35% and 0.11%, respectively. The radioactivity at the administration site was extremely high at all time points up to 144 h (>98% of injected dose). The small amount of radioactivity which did transfer from the administration site distributed mainly to the liver, spleen, kidney cortex and bone. Compared with the DW-166HC group, the group that received 166Ho nitrate alone displayed three- to fourfold higher levels of radioactivity in the main tissues, including liver, spleen, kidney cortex and bone, at 24 h after administration (P < 0.05).

CONCLUSION

The results of this study show clearly that most of the administered DW-166HC remained at the administration site. It is concluded that the chitosan complex may be used to retain 166Ho within a limited area in cancer of the prostate.

The interference of medical radionuclides with occupational in vivo gamma spectrometry

Radiation workers undergo routine monitoring for the evaluation of external and internal radiation exposures. The monitoring of internal exposures involves gamma spectrometry of the whole body (whole body counting) and measurements of excreta samples. Medical procedures involving internal administration of radioactive radionuclides are widely and commonly used. Medical radionuclides are typically short-lived, but high activities are generally administered, whereas occupational radionuclides are mostly long-lived and, if present, are found generally in relatively smaller quantities. The aim of the present work was to study the interference of some common medical radionuclides (201Tl, 9mTc, 57Co, and 131I) with the detection of internal occupational exposures to natural uranium and to 137Cs. Workers having undergone a medical procedure with one of the radionuclides mentioned above were asked to give frequent urine samples and to undergo whole body and thyroid counting with phoswich detectors operated at the Nuclear Research Center Negev. Urine and whole body counting monitoring were continued as long as radioactivity was detectable by gamma spectrometry. The results indicate that the activity of medical radionuclides may interfere with interpretation of occupational intakes for months after administration.

Assessment of intakes of mixtures of radionuclides

Australia has several uranium mines and a large number of mineral sand mines, with associated processing facilities. Exposures resulting from these mining and processing operations usually involve intakes of mixtures of radionuclides. This work describes the development of a suite of first order, linear compartment models, based on the ICRP Publication 66 respiratory tract model, and an analytical solution to the decay equations, for assessing the consequences of such intakes. The computer programs based on these models directly compute excretion, organ retention and organ and whole-body doses for intakes of either single radionuclides or any mixture of radionuclides belonging to the same radioactive decay chain. The intake can be via inhalation, ingestion or injection, and can be acute, chronic or of limited duration. The starting concentration and degree of secular (dis)equilibrium can be specified for each radionuclide. No assumptions need to be made about the relative magnitudes of the radioactive half-lives of the different nuclides.

Human biokinetics of inhaled terbium oxide

Four healthy men inhaled a monodisperse aerosol of 160Tb-labelled terbium oxide particles. The behaviour of the tracer was studied through measurements of body radioactivity and of its urinary and faecal excretion. Estimated early faecal losses in the four subjects ranged from 3% to 31% of the initial respiratory-tract deposit; most of the residue had become systemic within a year, with the principal deposit apparently in bone. Interference from this systemic deposit prevented accurate determination of the long-term pulmonary clearance kinetics, but the pattern was broadly what would be expected for Type M materials in the ICRP's Human Respiratory Tract Model. Averaged trends in the whole-body residue after approximately 1 year suggest a clearance half-life of approximately to 5 y.

Database of calculated values of retention and excretion for members of the public following acute intake of radionuclides

Intakes of radionuclides can be estimated from measurements of radioactivity in the whole body or in specific organs or in excreta by comparing them with predicted retention or excretion data calculated using standard biokinetic models. For occupational exposure monitoring, data are presented by ICRP for 29 radionuclides in Publication 78 (1997) and by the authors for 42 radionuclides as electronic look-up tables in Microsoft Excel. In the present work, values of retention and excretion were computed for selected radionuclides inhaled or ingested by members of the public. Graphs were constructed from the computed results showing the predicted monitoring data as functions of time following acute intakes of radionuclides. A graphical database was assembled on the Web site http//www.nirs.go.jp/RPD/ to provide a tool for the interpretation of bioassay measurements.

Eliminating bias in routine bioassay when there is an unknown time of intake

Routine bioassay programmes sometimes find evidence of an unsuspected intake. If there were no workplace indicators of exposure or intake, it is necessary to assume a value for the time of intake. Under these circumstances, the International Commission on Radiological Protection (ICRP) continues to recommend using the midpoint of the interval between routine bioassay measurements (ICRP Publication 78, paragraph 106). The assumption of T/2 as the time of intake, where T is the interval between bioassay measurements, represents the expectation value of the time of intake, (t), assuming uniform probability of an intake at any given time. This assumption results in a modest bias, of the expectation value of the intake, (I), that would have been received by a population of workers who had uniform probability over time of intake. This underestimation leads to a negative or positive bias in dose estimates derived in this fashion. The bias is characterised for realistic, routine urinalysis programs for Pu, U and 3H, as well as for in vivo measurements of 125I, 131I and 137Cs. Simple numerical methods are presented for correcting the bias. The bias is greatest for radionuclides whose half-lives are short with respect to the interval between bioassay measurements. Since the primary concern is estimating intake rather than time, the assumed time of intake should be chosen as t(I) rather than T/2. The ICRP should consider revising some of the tables in its Publication 78 to reflect this.

Guidance on individual monitoring programmes for radioisotopic techniques in molecular and cellular biology

The radioisotope techniques used in molecular and cellular biology involve external and internal irradiation risk. The personal dosemeter may be a reasonable indicator for external irradiation. However, it is necessary to control the possible internal contamination associated with the development of these techniques. The aim of this project is to analyse the most usual techniques and to establish programmes of internal monitoring for specific radionuclides (32P, 35S, 14C, 3H, 125I and 131I). To elaborate these programmes it was necessary to analyse the radioisotope techniques. Two models have been applied (NRPB and IAEA) to the more significant techniques, according to the physical and chemical nature of the radionuclides, their potential importance in occupational exposure and the possible injury to the genetic material of the cell. The results allowed the identification of the techniques with possible risk of internal contamination. It was necessary to identify groups of workers that require individual monitoring. The risk groups have been established among the professionals exposed, according to different parameters: the general characteristics of receptor, the radionuclides used (the same user can work with one, two or three radionuclides at the same time) and the results of the models applied. Also a control group was established. The study of possible intakes in these groups has been made by urinalysis and whole-body counter. The theoretical results are coherent with the experimental results. They have allowed guidance to individual monitoring to be proposed. Basically, the document shows: (1) the analysis of the radiosotopic techniques, taking into account the special containment equipment; (2) the establishment of the need of individual monitoring; and (3) the required frequency of measurements in a routine programme.

Developments in internal monitoring techniques

In an effort to increase accuracy and speed, improve detection limits and reduce uncertainties in internal dosimetry, laboratories have developed improved or new internal monitoring techniques in both in vivo measurements and bioassay analyses. Most of these techniques have not yet entered routine monitoring programmes. This paper intends to summarise these new techniques, show their potential improvements compared to the currently employed monitoring routines and discuss the main aspects of the EC-funded IDEA project, which aims at a comprehensive assessment of these techniques and the enhancements necessary to bring them to broader acceptance in the routine monitoring community.

Sensitivity analysis techniques applied to a revised model of molybdenum biokinetics

A revised model of molybdenum biokinetics in humans was recently developed on the basis of experimental data gathered in specific investigations conducted with stable tracers. The model can be used for radiation protection purposes, and it is also a suitable working tool for designing new investigations aimed at further improvements to the model. For the latter goal, a sensitivity analysis was performed in order to determine the most significant model parameters in relation to output measurements performed in studies of molybdenum metabolism. A typical sensitivity analysis approach was adopted, considering the effects in variation of model parameters on the time courses of model outputs such as urinary excretion and blood clearance. A recent new sensitivity technique was considered too, based on the calculation of the so-called generalised sensitivity functions. This combines the sensitivities of the model output with respect to model parameters (as in the typical sensitivity analysis method), with the sensitivities of parameter estimates with respect to changes in model outputs. The results obtained in this analysis suggests that data collected in the first 7 h are critical for the definition of the process of blood clearance and related parameters, whereas reliable information at later times is required for a proper characterisation of urinary excretion.

The excretion of biotrace elements using the multitracer technique in tumour-bearing mice

A radioactive multitracer solution obtained from the nuclear reaction of selenium with 25 MeV/nucleon 40Ar ions was used for investigation of trace element excretion into the faeces and urine of cancerous mice. The excretion rates of 22 elements (Na, K, Rb, Mg, Ca, Sr, Ga, As, Sc, V, Cr, Mn, Co, Fe, Y, Zr, Mo, Nb, Tc, Ru, Ag and In) were simultaneously measured under strictly identical experimental conditions, in order to clarify the excretion behavior of these elements in cancerous mice. The faecal and urinary excretion rates of Mg, Sr, Ga, As, Sc, V, Cr, Mn, Co, Fe, Y, Zr, Nb, Ru and Mo in cancerous mice, showed the in highest value at 0-8 hours. The accumulative excretion of Ca, Mo, Y and Zr was decreased and Na, Fe, Mn and Co increased in tumour-bearing mice, when compared to normal mice.

The effects of internal radiation exposure on cancer mortality in nuclear workers at Rocketdyne/Atomics International

We examined the effects of chronic exposure to radionuclides, primarily uranium and mixed-fission products, on cancer mortality in a retrospective cohort study of workers enrolled in the radiation-monitoring program of a nuclear research and development facility. Between 1950 and 1994, 2,297 workers were monitored for internal radiation exposures, and 441 workers died, 134 (30.4%) of them from cancer as the underlying cause. We calculated internal lung-dose estimates based on urinalysis and whole-body and lung counts reported for individual workers. We examined cancer mortality of workers exposed at different cumulative lung-dose levels using complete risk-set analysis for cohort data, adjusting for age, pay type, time since first radiation monitored, and external radiation. In addition, we examined the potential for confounding due to chemical exposures and smoking, explored whether external radiation exposure modifies the effects of internal exposure, and estimated effects after excluding exposures likely to have been unrelated to disease onset. Dose-response relations were observed for death from hemato- and lymphopoietic cancers and from upper aerodigestive tract cancers, adjusting for age, time since first monitored, pay type, and external (gamma) radiation dose. No association was found for other cancers, including cancers of the lung. Despite the small number of exposed deaths from specific cancer types and possible bias due to measurement error and confounding, the positive findings and strong dose-response gradients observed suggest carcinogenic effects of internal radiation to the upper aerodigestive tract and the blood and lymph system in this occupational cohort. However, causal inferences require replication of our results in other populations or confirmation with an extended follow-up of this cohort.

[Radiotoxicology]

Radiotoxicology is a science aiming firstly to estimate the biological effects induced by radiation in workers and general population after internal contamination of radionuclides, secondly evaluate the risk on health. After internal contamination, the analysis of biokinetics of radioactive compounds allow to understand their behaviour in the body. Those complex processes describe routes of radionuclide intake, direct blood uptake or transfer of soluble form to blood from deposit area, urine and fecal excretions, distribution and retention of radionuclides in different target organs. These processes are modelled to establish mathematical calculations. Data obtained are important to the interpretation of bioassay measurement for initial activity deposit expressed in becquerel (Bq: transformation.s-1) and committed effective dose calculation, expressed in sievert (Sv). This committed effective dose corresponds to the absorbed dose expressed in gray (Gy), weighted by a radiation weighting factor related to the quality of radiation and a tissue weighting factor which represents the contribution of the target organ to the total detriment due to effects induced by uniform irradiation of the whole body. This committed effective dose is a specific parameter for risk assessment which characterizes the radiotoxicology as a special part of toxicology.

Determination of aluminium-26 in biological materials by accelerator mass spectrometry

Studies of the biological chemistry of aluminium can gain significantly from the use of the long-lived isotope 26Al as a tracer, although the cost of the isotope often precludes its determination by radiochemical counting techniques. Accelerator mass spectrometry (AMS) provides an ultra-sensitive method of determination, free from isobaric interference from atomic (26Mg) or molecular species. The source materials for AMS can be aluminium oxide or phosphate, both of which can be readily prepared at a sufficient level of purity from biological substrates. Natural aluminium (27Al, 100%) is added to the preparations as a chemical yield monitor and to provide the reference for the isotope ratio measurement. 26Al/27Al ratios can be determined over the range 10(-14)-10(-7), implying a limit of detection for 26Al of around 10(-18) g. The precision of measurement and long-term reproducibility are < 5% and < 7% (RSD), respectively. Chemical methodologies for routine measurements on blood and urine samples have been developed.

Chlorine-36 in fossil rat urine: an archive of cosmogenic nuclide deposition during the past 40,000 years

Knowledge of the production history of cosmogenic nuclides, which is needed for geological and archaeological dating, has been uncertain. Measurements of chlorine-36/chlorine (36Cl/Cl) ratios in fossil packrat middens from Nevada that are radiocarbon-dated between about 38 thousand years ago (ka) and the present showed that 36Cl/Cl ratios were higher by a factor of about 2 before approximately 11 ka. This raises the possibility that cosmogenic production rates just before the close of the Pleistocene were up to 50% higher than is suggested by carbon-14 calibration data. The discrepancy could be explained by addition of low-carbon-14 carbon dioxide to the atmosphere during that period, which would have depressed atmospheric radiocarbon activity. Alternatively, climatic effects on 36Cl deposition may have enhanced the 36Cl/Cl ratios.

205Bi/206Bi cyclotron production from Pb-isotopes for absorption studies in humans

Pb(p, xn) thick target excitation functions were measured in the energy range 10-38 MeV in order to optimize the production of isotopically pure radiobismuth from natPb, 206Pb, and 207Pb. Additionally, the decay of Po-isotopes from deuteron irradiation of natural bismuth (209Bi) was exploited for radiobismuth production. 205Bi was produced from 206Pb at 20 MeV with only 2% of 206Bi at 4 weeks post irradiation. Bismuth compounds as used in the treatment of peptic ulcer were labeled with 205Bi for absorption studies in animals and subjects.

Analysis of urine samples from metastatic bone cancer patients administered 153Sm-EDTMP

153Sm-EDTMP is currently undergoing clinical evaluation as a radiotherapeutic agent for the relief of pain associated with cancer metastatic to bone. These clinical studies have demonstrated biodistributions similar to those seen earlier in animals, namely, rapid clearance from blood, selective uptake in bone and in particular metastatic bone lesions. The radioactivity not deposited in bone is cleared through the kidneys into the urine. In this study, urine samples collected from 9 patients injected with 153Sm-EDTMP underwent complexation analysis via Pharmacia SP-Sephadex C25 cation exchange chromatography. The results showed 96.9 +/- 1.7% of the radioactivity in the urine to be present as a complex of 153Sm. An HPLC method was developed and it was demonstrated that different complexes of 153Sm could be separated. A non-radioactive analytical standard of the Sm-EDTMP chelate was synthesized, characterized and shown to have the same HPLC retention profile as the 153Sm-EDTMP drug product. HPLC analysis was performed on six urine samples and in each case a single radioactivity peak with an elution profile the same as that of a 153Sm-EDTMP standard was observed. These results indicate that the 153Sm-EDTMP chelate is excreted intact in the urine of patients.

Pharmacokinetic analysis of blood distribution of intravenously administered 153Gd-labeled Gd(DTPA)2- and 99mTc(DTPA) in rats

Rat plasma distribution data obtained following IV administration of 99mTc(DTPA) alone or after co-administration of 99mTc(DTPA) and 153Gd-labeled Gd(DTPA)2- at 0.001, 0.1, and 1.0 mmol Gd/kg were evaluated using compartmental modeling techniques. A three-compartment open model was found to fit the data significantly better (P less than 0.01) than a two- or four-compartment open model. This model incorporates and links the plasma and urine data and includes a delay to account for the transit time through the kidneys/ureters. The two nonplasma compartments of the model were assumed to be related to rapidly and slowly equilibrating tissues. Tc(DTPA) and Gd(DTPA)2- had nearly identical pharmacokinetic profiles in plasma and the rate constants were essentially the same. No significant dose dependent pharmacokinetic differences were found for the range of Gd(DTPA)2- doses tested. Simulations of the proposed three-compartment model were used to generate concentration-time curves for each of the three compartments.

The role of Poison Control Centers in radiation accidents

In the days after the nuclear reactor accident at Chernobyl (USSR) in April 1986, the Dutch Poison Control Center had to answer questions concerning possible health effects caused by (over)exposure to ionizing radiation. These questions were similar to questions asked regarding exposure to toxic agents after chemical accidents. It is obvious that the experience and practical approach of a Poison Control Center in handling toxicological problems can be used in problems concerning ionizing radiation.

Metabolism of 51Cr, 54Mn, 59Fe and 60Co in lactating dairy cows

Different radionuclides which can be released by nuclear installations (Na2(51) CrO4 , 54MnCl2 , 59FeCl3 and 60CoCl2 ) were given to lactating dairy cows either orally or by intravenous (i.v.) injection. Excretion into feces and urine as well as secretion into milk were followed for several weeks. Distribution of activity in organs was determined at sacrifice 102 days after oral and 70 days after i.v. administration. After ingestion, excretion of chromate followed a three-term exponential function with half-lives of 0.88, 3.7 and 26 days. Intestinal absorption was on the order of 0.1-0.2%. About 63% of injected chromate was excreted into urine, about 18% into feces and about 3.6% into milk. Orally administered chromate was concentrated in liver, intestine and spleen. The transfer coefficient into milk was about 1 X 10(-5) days/1. Less than 1% of an ingested dose of manganese was absorbed. Excretion occurred mainly into feces and followed a three-term exponential function. Very little manganese was excreted into urine or secreted into milk. Manganese is concentrated in brain, pancreas, kidney and heart but the differences in concentration are small. The body burden is, therefore, mainly determined by manganese in muscle, skin and bone. The transfer coefficient of manganese into milk is about 3 X 10(-5). Excretion of iron into feces after oral administration follows a three-term exponential function with a small component having a half-life of 72 days. Intestinal absorption is on the order of 0.5-2% of the dose. After i.v. administration, 7% of the dose is excreted in the feces displaying two components of turnover. Very little iron is excreted into urine. Secretion into milk also follows a two-term exponential function. The transfer coefficient of iron into milk is about 3 X 10(-5). About 0.6% of an oral and about 94% of an i.v. dose were recovered from the cows at sacrifice. Most activity was present in blood, liver and spleen. Excretion of radiocobalt into feces after oral administration is described by a three-term exponential function, while excretion into milk and urine is described by two-term exponential functions. Long-lived components in urine and milk represent about 5-10% of the activity absorbed. Only 1-2% of an oral cobalt dose is absorbed and 0.05% of an oral and 5.85% of an i.v. dose is recovered from the cow at sacrifice where concentrations are highest in kidney and thyroid. The body burden depends mainly on cobalt in muscle, skin and bone. The transfer coefficient of cobalt into milk is about 7.5 X 10(-5).

The development and application of alpha-particle spectrometry for low-level-radioactivity measurements in biological and environmental materials

The design, construction and use of a purpose-built alpha-particle spectrometer is presented. The application of the spectrometer in the determination of low levels of radioactivity in biological and environmental materials is also discussed.

The impact of natural radioactivity from a coal-fired power plant

In a coal-fired power station burning coal which contained between 14--100 ppm U, 210Pb was detected in the urine of an exposed group of individuals. Chromosome aberrations (complex, numerical and the percentage of total aberrations) were also registered.

The effect of certain variables on the tumor and tissue distribution of tracers. Part 1; Carrier

Buffalo rats bearing thigh-implanted strain-7777 Morris hepatomas were used as a model for studying the effect of carrier material on the body distribution, tumor uptake, excretion, and tumor-to-background ratios of 67Ga and 54Mn. An effort was also made to observe the changes in 67Ga and 54Mn concentrations induced by carrier in viable tumor and skeletal muscle, relative to their interstitial fluid space. This value is referred to as the Tissue Distribution Index. Carrier manipulation resulted in striking changes in the distribution of the two ions from the carrier-free state. The data also indicated a difference in the pharmacodynamics of 67Ga and 54Mn in malignant and healthy tissues which could be of importance to nuclear medicine and oncology.

Blood and urinary 18F pharmacokinetics following parenteral administration in the rat

Blood and urinary excretion time courses of 18F administered parenterally to rats were monitored for two hours. The intraperitoneal, subcutaneous, and intravenous routes gave kinetically indistinguishable results after ten minutes following the dose. The blood time course during the first hour following intramuscular dosing showed a relative constancy and suggested a delayed absorption time.

Concepts and limitations in the application of radiolabeled antiandrogens, estrogens, or androgens as isotopic scanning agents for the prostate

The distribution of labeled antiandrogens, estrogens, and androgens has been studied in the rat and dog to elucidate the principles and limitations encountered when these agents are used in radioscanning the prostate. The bulk of the label is rapidly cleared in the biliary secretions and in the urine. The concentration of label in the prostate in comparison to other tissues such as the intestines limits the application of these agents in radioscanning of the prostate. The tissue distributions were confirmed by total body radioscans of rats and dogs receiving a labeled 3'-iodinated analog of the antiandrogen flutamide (Sch 13521 or 3'-trifluoromethyl-4'-nitroisobutyranilide).