Radioactivity measurements of former military personnel exposed to weapon debris.

Sixteen former military personnel who were present at the "Smoky" atmospheric nuclear weapon test have been investigated for internal deposits of radioactivity. Whole-body and thorax gamma-ray measurements, thorax and skeletal actinide measurements, and urinalyses for plutonium-239 and strontium-90 indicated no evidence of radioactivity in excess of that found in the general population.

Retention of plutonium in the beagle after gastrointestinal absorption.

A 0.01 M bicarbonate solution containing 130 nCi (approximately 5 kBq) of 237Pu (90% hexavalent, 93% ultrafilterable) was administered via gelatin capsule to six adult male beagles following a 21-hr fast. The dogs were sacrificed after 5-6 weeks and the percentage of the administered plutonium retained in the liver plus skeleton and its distribution within the skeleton were determined. The mean amount retained in these tissues was (0.063 +/- 0.006)% of the administered dose. The mean amounts of plutonium in the liver and the skeleton were approximately equal, and the distribution within the skeleton was similar to that observed by other workers following either intravenous injection or inhalation. Our value for plutonium retention by the dog is about a factor of three less than the values we have reported for rodents. It is a factor of 7 greater than the product of the values for gastrointestinal absorption (f1) and fractional retention (f2) recommended by the ICRP for man.

Correlation of lung dose with Rn concentration, potential alpha-energy concentration and daughter surface deposition: a Monte Carlo analysis.

In the evaluation of lung cancer risk from Rn daughters in the home environment, measurements of Rn concentration or potential alpha-energy concentration (PAEC) have served as proxies for actual radiation dose to the lungs. This paper uses a single-compartment room model, model parameters from a number of recent studies, and Monte Carlo analysis to explore the relative value of Rn concentration, PAEC, and room surface deposition, or plate-out, as estimators of lung dose. Results indicate that Rn concentration and PAEC are fairly good estimators, explaining roughly 50-70% of the variation in dose for the conditions studied, over the range of Rn concentrations considered (74-555 Bq m 3). The relative advantage of one measure over the other depends upon the variability in a number of factors across the houses being evaluated. Room surface activity deposition was not found to be superior to the traditional measures of Rn concentration and PAEC. Significant room for improvement remains in the development of simple home monitors providing an improved estimation of lung dose.

Overview and dosimetry of the Hanford Americium accident case.

Postmortem examinations of selected tissues from an individual who received a massive internal exposure to 241Am and succumbed to pre-existing cardiovascular disease 11 y later are summarized. Significant findings include acellularity of the marrow, marked peritrabecular fibrosis, and a lack of bone surface remodeling, confirmed by bone-surface alpha-spectrometry; tissue concentrations of 241Am that indicate bone and liver as the primary deposition sites, in general agreement with the new ICRP model, but not with the older models; a distribution of 241Am in soft tissues in general agreement with that observed in animals; and the absence of any other gross or microscopic pathological findings attributable to the exposure. Cumulative absorbed doses to the bone, bone surface, liver, and lung were 18, 520, 8, and 1.6 Gy, respectively. The probability of not observing a fatal cancer based on BEIR-IV risk factors for these absorbed doses was 12%.

Measurement techniques for radium and the actinides in man at the Center for Human Radiobiology.

Various techniques are employed to determine the amounts, retention, and distribution of radioactivity in human subjects in vivo. The principal method is gamma-ray spectrometry with large NaI(Tl) scintillation crystals ("whole-body counting"). The geometries used include an arc of 1.5-m radius in which all parts of the body are roughly equidistant from the detector, a reclining chair and a flat bed with detectors placed above and below the subject. When a sufficient amount of radioactivity is present in a subject, scanning techniques assist in determining its distribution in the body. Specialized instruments such as a xenon-filled proportional counter and a dual-crystal (phoswich) detector are used to measure low-energy photon emitters, primarily plutonium and americium. americium. There are three primary methods of calibrating the detectors. The first is analytical, in which a rigorous mathematical treatment is employed; the second involves the administration of tracer amounts of radioactivity to human volunteers; the third consists of determining detector response to known amounts of radioactivity in a phantom. All three methods can be intercompared, and further evaluated by comparing the results of measurements in vivo with those of postmortem analyses. For both radium and thorium cases measured in vivo, the interpretation of the results is complicated by the fact that neither radium nor thorium emit gamma rays of any consequence. Instead, the observed gamma rays result from the decay of 214Bi (RaC) and 208Tl (ThC"), respectively. Since each of these nuclides is preceded in the decay chain by an isotope of the noble gas radon, some of which is exhaled, its activity is not equal to that of the parent radium or thorium. Therefore, breath samples are collected to determine the exhalation rate of the precursor isotope, 222Rn (radon) or 220Rn (thoron). The total body content is then the sum of the gamma activity and the exhaled radioactivity, referred to as the retained and emanating fractions, respectively.

Postmortem tissue contents of 241Am in a person with a massive acute exposure.

241Am was determined radiochemically in the tissues of USTUR Case 246, a 76-y-old man who died of cardiovascular disease 11 y after massive percutaneous exposure following a chemical explosion in a glove box. This worker was treated extensively with a chelation drug, DTPA, for over 4 y after exposure. The estimated 241Am deposition at the time of death was 540 kBq, of which 90% was in the skeleton, 5.1% in the liver, and 3.5% in muscle and fat. Among the soft tissues, the highest concentrations were observed in liver (22 Bq g-1), certain cartilaginous structures such as the larynx (15 Bq g-1) and the red marrow (9.7 Bq g-1), as compared with the mean soft tissue concentration of approximately 1 Bq g-1. Concentration in muscle was approximately that of the soft tissue average, while concentrations in the pancreas, a hilar lymph node and fat were less than the average. Concentrations in bone ash were inversely related to the ratio of ash weight to wet weight, a surrogate for bone volume-to-surface ratio. The distribution of activity in this case is reasonably consistent with that observed in another human case, when allowance is made for chelation therapy, and also tends to support more recent models of 241Am metabolism.

Bone surface concentrations and dose rates 11 years after massive accidental exposure to 241Am.

Alpha-particle spectrograms of bone samples from USTUR Case 246 were analyzed to determine the depth in tissue from which the 241Am alpha-particles were emitted. In four samples of bone, the lack of energy straggling in the alpha spectra indicated that essentially all the 241Am was deposited directly on the exposed bone surface, and none had been translocated to within bone volume. These findings agree well with the results of autoradiographic examination of bone samples from the same case, but are in marked contrast to findings on another case (USTUR Case 102) who had been exposed to a much smaller amount of 241Am at a younger age and survived approximately twice as long after the exposure. It is problematical whether the lack of 241Am redistribution, and therefore the implied absence of bone remodeling, in Case 246 was due to his advanced age at exposure or to a deterministic effect of alpha-irradiation on bone metabolism, but the observation of radiation effects on bone metabolism in former radium workers supports the latter.

Deterministic effects of 241Am exposure in the Hanford Americium accident case.

Lifetime follow-up of USTUR Case 246 demonstrated the lack of severe biological effects resulting from his exposure to 241Am. Deterministic effects observed were limited to hematological changes, including lymphopenia and thrombocytopenia. These hematological changes were consistent with those observed in experimental animals following actinide exposure. Cataracts were removed from the left and right eyes at 547 and 1,030 d after the accident, respectively, but were considered to be trauma-induced rather than radiation-induced. No abnormal findings were reported from gross or histological examinations of tissue samples removed at autopsy, other than those resulting from the subject's preexisting cardiovascular disease.

The effects of sera from Ra dial painters on the mitogenic responses of normal human lymphocytes.

Sera from former Ra workers were incubated with normal human lymphocytes and the effects produced in stimulation assays with specific mitogens were compared with those observed with incubation using normal human sera. The sera from the Ra workers with high residual body burdens (greater than 0.1 muCi 226Ra) inhibited normal control lymphocytes stimulated with Con A and PHA when compared with control homologous sera. Although a significant suppression was not obtained between high- and low-body-burden groups a trend was indicated. Sera from the highest body-burden worker (0.58 muCi) was inhibitory to all normal lymphocyte populations tested regardless of the mitogen stimulant employed.

Excretion of depleted uranium by Gulf War veterans.

During the Persian Gulf War, in 1991, approximately 100 US military personnel had potential intakes of depleted uranium (DU), including shrapnel wounds. In 1993, the US government initiated a follow-up study of 33 Gulf War veterans who had been exposed to DU, many of whom contained embedded fragments of DU shrapnel in their bodies. The veterans underwent medical evaluation, whole-body counting, and urinalysis for uranium by kinetic phosphorescence analysis (KPA). Data are available from seven individuals who exceeded the detection limit for whole-body counting and also had elevated urinary uranium. Urinary excretion rates, in microg U g(-1) creatinine, were determined in 1997 and 1999. The body contents, in mg DU, were determined in 1997; it is assumed there were no significant decreases in total body content in the interim. For the 1997 data, the mean fractional excretion was (2.4 +/- 2.8) x 10(-5) g(-1) creatinine, and for the 1999 data, the mean was (1.1 +/- 0.6) x 10(-5) g(-1) creatinine. However, these means are not significantly different, nor is there any correlation of excretion rate with body content. Thus, human data available to date do not provide any basis for determining the effects of particle surface area, composition and solubility, and biological processes such as encapsulation, on the excretion rate.

Internal dose assessment in radiation accidents.

Although numerous models have been developed for occupational and medical internal dosimetry, they may not be applicable to an accident situation. Published dose coefficients relate effective dose to intake, but if acute deterministic effects are possible, effective dose is not a useful parameter. Consequently, dose rates to the organs of interest need to be computed from first principles. Standard bioassay methods may be used to assess body contents, but, again, the standard models for bioassay interpretation may not be applicable because of the circumstances of the accident and the prompt initiation of decorporation therapy. Examples of modifications to the standard methodologies include adjustment of biological half-times under therapy, such as in the Goiania accident, and the same effect, complicated by continued input from contaminated wounds, in the Hanford 241Am accident.

Dose coefficients for intakes of radionuclides via contaminated wounds.

The NCRP Wound Model, which describes the retention of selected radionuclides at the site of a contaminated wound and their uptake into the transfer compartment, has been combined with the ICRP element-specific systemic models for those radionuclides to derive dose coefficients for intakes via contaminated wounds. These coefficients can be used to generate derived regulatory guidance (i.e., the activity in a wound that would result in an effective dose of 20 or 50 mSv, or in some cases, a organ-equivalent dose of 500 mSv) and clinical decision guidance (i.e., activity levels that would indicate the need for consideration of medical intervention to remove activity from the wound site, administration of decorporation therapy or both). Data are provided for 38 radionuclides commonly encountered in various activities such as nuclear weapons, fuel fabrication or recycling, waste disposal, medicine, research, and nuclear power. These include 3H, 14C, 32P, 35S, 59Fe, 57,58,60Co, 85,89,90Sr, 99mTc, 106Ru, 125,129,131I, 134,137Cs, 192Ir, 201Tl, 210Po, 226,228Ra, 228,230,232Th, 234,235,238U, 237Np, 238,239,240,241Pu, 241Am, 242,244Cm, and 252Cf.

The AAPM/RSNA physics tutorial for residents: internal radiation dosimetry: principles and applications.

Internal dose calculations in nuclear medicine normally use the techniques, equations, and resources provided by the Medical Internal Radiation Dose (MIRD) Committee of the Society of Nuclear Medicine. The MIRD schema uses a unique set of symbols and quantities to calculate the absorbed dose of radiation in any target organ per radioactive decay in any source organ. The calculations involve the energy emitted per radioactive decay, the fraction of the emitted energy that is absorbed in various target organs, the masses of these organs, and both the physical decay and biologic clearance of the injected radioactive material. Standardized mathematical models (phantoms) of the human body and standardized biokinetic models are also used. A computer program, MIRDose, calculates dose tables per unit administered activity of various radiopharmaceuticals. Special care must be taken when nuclear medicine procedures involve pregnant or lactating patients. New methodologies are becoming available to calculate doses to individual patients.