[en] Following a radioactive dispersal device (RDD) incident, it may be necessary to evaluate the internal contamination levels of a large number of potentially affected individuals to determine if immediate medical follow-up is necessary. Since the current laboratory capacity to screen for internal contamination is limited, rapid field screening methods can be useful in prioritizing individuals. This study evaluated the suitability of a radiation portal monitor for such screening. A model of the portal monitor was created for use with models of six anthropomorphic phantoms in Monte Carlo N-Particle Transport Code Version 5 (MCNP) X-5 Monte Carlo Team (MCNP A General Monte Carlo N-Particle Transport Code Version 5. LA-CP-03-0245. Vol. 2. Los Alamos National Laboratory, 2004.). The count rates of the portal monitor were simulated for inhalation and ingestion of likely radionuclides from an RDD for each of the phantoms. The time-dependant organ concentrations of the radionuclides were determined using Dose and Risk Calculation Software Eckerman, Leggett, Cristy, Nelson, Ryman, Sjoreen and Ward (Dose and Risk Calculation Software Ver. 8.4. ORNL/TM-2001/190. Oak Ridge National Laboratory, 2006.). Portal monitor count rates corresponding to a committed effective dose E(50) of 10 mSv are reported.

[en] Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10^-9 to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations.

[en] The personal dose equivalent, H_p(d), is the quantity recommended by the International Commission on Radiation Units and Measurements (ICRU) to be used as an approximation of the protection quantity Effective Dose when performing personal dosemeter calibrations. The personal dose equivalent can be defined for any location and depth within the body. Typically, the location of interest is the trunk where personal dosemeters are usually worn and in this instance a suitable approximation is a 30 cm X 30 cm X 15 cm slab-type phantom. For this condition the personal dose equivalent is denoted as H_p,slab(d) and the depths, d, are taken to be 0.007 cm for non-penetrating and 1 cm for penetrating radiation. In operational radiation protection a third depth, 0.3 cm, is used to approximate the dose to the lens of the eye. A number of conversion coefficients for photons are available for incident energies up to several MeV, however, data to higher energies are limited. In this work conversion coefficients up to 1 GeV have been calculated for H_p,slab(10) and H_p,slab(3) using both the kerma approximation and by tracking secondary charged particles. For H_p(0.07) the conversion coefficients were calculated, but only to 10 MeV due to computational limitations. Additionally, conversions from air kerma to H_p,slab(d) have been determined and are reported. The conversion coefficients were determined for discrete incident energies, but analytical fits of the coefficients over the energy range are provided. Since the inclusion of air can influence the production of secondary charged particles incident on the face of the phantom conversion coefficients have been determined both in vacuo and with the source and slab immersed within a sphere in air. The conversion coefficients for the personal dose equivalent are compared to the appropriate protection quantity, calculated according to the recommendations of the latest International Commission on Radiological Protection (ICRP) guidance.

[en] After a radiological dispersal device (RDD) event, people could become internally contaminated by inhaling dispersed radioactive particles. A rapid method to screen individuals who are internally contaminated is desirable. Such initial screening can help in prompt identification of those who are highly contaminated and in prioritizing individuals for further and more definitive evaluation such as laboratory testing. The use of handheld plastic scintillators to rapidly screen those exposed to an RDD with gamma-emitting radionuclides was investigated in this study. The Monte Carlo N-Particle transport code was used to model two commercially available plastic scintillation detectors in conjunction with anthropomorphic phantom models to determine the detector response to inhaled radionuclides. Biokinetic models were used to simulate an inhaled radionuclide and its progression through the anthropomorphic phantoms up to 30 d after intake. The objective of the study was to see if internal contamination levels equivalent to 250 mSv committed effective dose equivalent could be detected using these instruments. Five radionuclides were examined: ⁶⁰Co, ¹³⁷Cs, ¹⁹²Ir, ¹³¹I and ²⁴¹Am. The results demonstrate that all of the radionuclides except ²⁴¹Am could be detected when placing either one of the two plastic scintillator detector systems on the posterior right torso of the contaminated individuals.

[en] For internal dose calculations for the Mayak worker epidemiological study, quantitative estimates of uncertainty of the urine measurements are necessary. Some of the data consist of measurements of 24h urine excretion on successive days (e.g. 3 or 4 days). In a recent publication, dose calculations were done where the uncertainty of the urine measurements was estimated starting from the statistical standard deviation of these replicate mesurements. This approach is straightforward and accurate when the number of replicate measurements is large, however, a Monte Carlo study showed it to be problematic for the actual number of replicate measurements (median from 3 to 4). Also, it is sometimes important to characterize the uncertainty of a single urine measurement. Therefore this alternate method has been developed. A method of parameterizing the uncertainty of Mayak urine bioassay measmements is described. The Poisson lognormal model is assumed and data from 63 cases (1099 urine measurements in all) are used to empirically determine the lognormal normalization uncertainty, given the measurement uncertainties obtained from count quantities. The natural logarithm of the geometric standard deviation of the normalization uncertainty is found to be in the range 0.31 to 0.35 including a measurement component estimated to be 0.2.

[en] Accelerator-produced radiation levels at the perimeter of the Ernest Orlando Lawrence Berkeley National Laboratory (the Berkeley Laboratory) reached a maximum in 1959. Neutrons produced by the Bevatron were the dominant component of the radiation field. Radiation levels were estimated from measurements of total neutron fluence and reported in units of dose equivalent (rem). Accurate conversion from total fluence to dose equivalent demands knowledge of both the energy spectrum of accelerator-produced neutrons and the appropriate conversion coefficient functions for different irradiation geometries. At that time (circa 1960), such information was limited, and it was necessary to use judgment in the interpretation of measured data. The Health Physics Group of the Berkeley Laboratory used the best data then available and, as a matter of policy, reported the most conservative (largest) values of dose equivalent supported by their data. Since the early sixties, significant improvements in the information required to compute dose equivalent, particularly in the case of conversion coefficients, have been reported in the scientific literature. This paper reinterprets the older neutron measurements using the best conversion coefficient data available today. It is concluded that the dose equivalents reported in the early sixties would be reduced by at least a factor of two using current methods of analysis

[en] Dose-response data are presented from a large percentage of the US workers who were exposed to radium through the painting of luminous dials. The data in this paper are only from females, because very few males worked in this occupation. Log-normal analyses were done for radium-induced bone sarcomas and head carcinomas after the populations of the respective doses were first determined to be log-normally distributed. These populations included luminisers who expressed no radium-related cancerous condition. In this study of the female radium luminisers, the most important data concerning radiation protection are probably from workers who were exposed to radium but showed no cancer incidence. A total of 1391 subjects with average measured skeletal doses below 10 Gy are in this category. A primary purpose is to illustrate the strong case that ^226,228Ra is representative of those radionuclides that exemplify in humans a 'threshold' dose, a dose below which there has been no observed health effects on the exposed individual. Application of a threshold dose for radium deposited in the skeleton does not mean to imply that any other source of skeletal irradiation should be considered to follow a similar pattern. Second, a policy issue that begs for attention is the economic consequence of forcing radiation to appear as a highly toxic insult. It is time to evaluate the data objectively instead of formatting the extrapolation scheme beforehand and forcing the data to fit a preconceived pattern such as linearity through the dose-effect origin. In addition, it is time to re-evaluate (again) variations in background radiation levels throughout the world and to cease being concerned with, and regulating against, miniscule doses for which no biomedical effects on humans have ever been satisfactorily identified or quantified. (author)

No abstract available

[en] The impact a revision of nuclear decay data had on dose coefficients was studied using data newly published in ICRP Publication 107 (ICRP 107) and existing data from ICRP Publication 38 (ICRP 38). Committed effective dose coefficients for occupational inhalation of radionuclides were calculated using two sets of decay data with the dose and risk calculation software DCAL for 90 elements, 774 nuclides and 1572 cases. The dose coefficients based on ICRP 107 increased by over 10% compared with those based on ICRP 38 in 98 cases, and decreased by over 10% in 54 cases. It was found that the differences in dose coefficients mainly originated from changes in the radiation energy emitted per nuclear transformation. In addition, revisions of the half-lives, radiation types and decay modes also resulted in changes in the dose coefficients.

[en] The calculation of absorbed dose in skeletal tissues at radiogenic risk has been a difficult problem because the relevant structures cannot be represented in conventional geometric terms nor can they be visualized in the tomographic image data used to define the computational models of the human body. The active marrow, the tissue of concern in leukemia induction, is present within the spongiosa regions of trabecular bone, whereas the osteoprogenitor cells at risk for bone cancer induction are considered to be within the soft tissues adjacent to the mineral surfaces. The International Commission on Radiological Protection (ICRP) recommends averaging the absorbed energy over the active marrow within the spongiosa and over the soft tissues within 10 mm of the mineral surface for leukemia and bone cancer induction, respectively. In its forthcoming recommendation, it is expected that the latter guidance will be changed to include soft tissues within 50 mm of the mineral surfaces. To address the computational problems, the skeleton of the proposed ICRP reference computational phantom has been subdivided to identify those voxels associated with cortical shell, spongiosa and the medullary cavity of the long bones. It is further proposed that the Monte Carlo calculations with these phantoms compute the energy deposition in the skeletal target tissues as the product of the particle fluence in the skeletal subdivisions and applicable fluence-to-dose-response functions. This paper outlines the development of such response functions for photons