Evaluation Of Equivalent Dose Research Articles

Evaluation of equivalent dose from neutrons and activation products from a 15-MV X-ray LINAC.

A high-energy photon beam that is more than 10 MV can produce neutron contamination. Neutrons are generated by the [γ,n] reactions with a high-Z target material. The equivalent neutron dose and gamma dose from activation products have been estimated in a LINAC equipped with a 15-MV photon beam. A Monte Carlo simulation code was employed for neutron and photon dosimetry due to mixed beam. The neutron dose was also experimentally measured using the Optically Stimulated Luminescence (OSL) under various conditions to compare with the simulation. The activation products were measured by gamma spectrometer system. The average neutron energy was calculated to be 0.25 MeV. The equivalent neutron dose at the isocenter obtained from OSL measurement and MC calculation was 5.39 and 3.44 mSv/Gy, respectively. A gamma dose rate of 4.14 µSv/h was observed as a result of activations by neutron inside the treatment machine. The gamma spectrum analysis showed 28Al, 24Na, 54Mn and 60Co. The results confirm that neutrons and gamma rays are generated, and gamma rays remain inside the treatment room after the termination of X-ray irradiation. The source of neutrons is the product of the [γ,n] reactions in the machine head, whereas gamma rays are produced from the [n,γ] reactions (i.e. neutron activation) with materials inside the treatment room. The most activated nuclide is 28Al, which has a half life of 2.245 min. In practice, it is recommended that staff should wait for a few minutes (several 28Al half-lives) before entering the treatment room after the treatment finishes to minimize the dose received.

Shielding calculation for the Proton-Therapy-Center in Prague, Czech Republic

The study of radiation fields around the cyclotron room was carried out for the Proton-Therapy-Center in Prague (Czech Republic). The 230 MeV proton cyclotron will be installed there as well as additional components located along the beampath. The ambient doses were computed by taking into account 3 major sources of secondary radiation (neutron and photon) – the cyclotron, energy degrader and its collimator. The fluxes of neutrons/photons produced by protons were computed using Monte Carlo code MCNPXTM 2.5.0. Energy and angular distributions were computed for various targets and proton beam energies. Different physical models of high-energy proton interactions were studied as well. The annual charge at the cyclotron exit was based on the “patient model” (distribution of treatment type) established for a similar therapy center in Essen (Germany). Calculated data were used as an approximation of the radiation field of the most significant sources of radiation (graphite degrader, tantalum collimator). Fluence-to-dose conversion factors from ICRP-74 were used for the annual dose equivalent evaluation. With these results, the shielding analysis in the Therapy-Center was performed in the context of the current legislation requirements.

Importance of personal dose equivalent evaluation in Fukushima in overcoming social panic

The relationship between the reported ambient dose equivalent (H*(10)) and the individual dose rate recorded by medical staff in Fukushima City after the accident at the Fukushima Daiichi nuclear power plant was evaluated, following a 9.0-magnitude earthquake that struck the east coast of Japan. Personal dose equivalent (H(p)(10)) ranged from 0.08 to 1.63 µSv h(-1) and H*(10) ranged from 0.86 to 12.34 µSv h(-1). H(p)(10) from March to July 2011 were significantly lower than H*(10). The relationships between these dose equivalents were moderately correlated. The regression equation was calculated as follows: H(p)(10)=0.0696×H*(10)+0.0538. The preliminary data of this study show that, in Fukushima, the individual dose is much lower than that determined H*(10). It is important to evaluate H(p)(10) in order to lessen the anxiety of the general population in Fukushima.

Dose evaluation from multiple detector outputs using convex optimisation

A dose evaluation using multiple radiation detectors can be improved by the convex optimisation method. It enables flexible dose evaluation corresponding to the actual radiation energy spectrum. An application to the neutron ambient dose equivalent evaluation is investigated using a mixed-gas proportional counter. The convex derives the certain neutron ambient dose with certain width corresponding to the true neutron energy spectrum. The range of the evaluated dose is comparable to the error of conventional neutron dose measurement equipments. An application to the neutron individual dose equivalent measurement is also investigated. Convexes of particular dosemeter combinations evaluate the individual dose equivalent better than the dose evaluation of a single dosemeter. The combinations of dosemeters with high orthogonality of their response characteristics tend to provide a good suitability for dose evaluation.

The neutron dose equivalent evaluation and shielding at the maze entrance of a Varian Clinac 23EX treatment room

To evaluate the neutron and photon dose equivalent rate (H(n,D) and H(G)) at the outer maze entrance and the adjacent treatment console area after the installation of a Varian Clinac 23EX accelerator with a higher beam energy than its predecessor. The evaluation was based on measurements and comparison with several empirical calculations. The effectiveness of borated polyethylene (BPE) boards, as a maze wall lining material, on neutron dose and photon dose reduction is also reported. A single energy Varian 6 MV photon linear accelerator (linac) was replaced with a Varian Clinac 23EX accelerator capable of producing 18 MV photons in a vault originally designed for the former accelerator. In order to evaluate and redesign the shielding of the vault, the neutron dose equivalent H(n,D) was measured using an Andersson-Braun neutron Rem meter and the photon dose equivalent HG was measured using a Geiger Müller and an ion chamber gamma-ray survey meter at the outer maze entrance. The measurement data were compared to semiempirical calculations such as the Kersey method, the modified Kersey method, and a newly proposed method by Falcão et al. Additional measurements were taken after BPE boards were installed on the maze walls as a neutron absorption lining material. With the gantry head tilted close to the inner maze entrance and with the jaws closed, both neutron dose equivalent and photon dose equivalent reached their maximum. Compared to the measurement results, the Kersey method overestimates the neutron dose equivalent H(n,D) by about two to four times (calculation/measurement ratio approximately 2.4-3.8). Falcão's method largely overestimates the H(n,D) (calculation/measurement ratio approximately 3.9-5.5). The modified Kersey method has a calculation to measurement ratio about 0.6-0.9. The photon dose equivalent calculation including McGinley's capture gamma dose equivalent equation estimates about 77%-98% of the measurement. After applying BPE boards as a lining material on the inner corner of the maze wall, the H(n,D) and the H(G) at maze entrance were decreased by 41% and 59%, respectively. This work indicates that the Kersey method overestimates the neutron dose equivalent H(n,D) for a Varian Clinac 23EX accelerator. The Falcão method overestimates the H(n,D) partially due to the discrepancy in the International Commission on Radiological Protection (ICRP) conversion factors caused by the uncertainties of the estimated average neutron energy. The modified Kersey method gives the closest estimation of a Varian Clinac 23EX accelerator operated at 18 MV photon mode in a maze with a similar design as in the authors' study. However, it should be used with caution because of its tendency to underestimate the H(n,D). A borated polyethylene lining can provide a cost effective method to reduce neutron and photon dose equivalent at the maze door for an existing linac vault, following the installation of a higher energy linac.

TL-efficiency—Overview and experimental results over the years

For several decades, thermoluminescence detectors (TLDs) have been widely used in environmental, personal and medical dosimetry. With the development of heavy-ion accelerators, investigations concentrated on the thermoluminescent (TL) response of different TL phosphors to heavy charged particles (HCPs), motivated by the manifold potential TLD applications in HCP dosimetry. At the beginning, the research interest was focused particularly on the most common TL material, LiF:Mg, Ti. During the last years, considerable effort has been invested to determine precisely TL-efficiency also for other phosphors, either based on theoretical models or on experiments at heavy-ion accelerator facilities. The high amount of TLDs being used in space dosimetry in combination with plastic nuclear track detectors (PNTDs) essentially requires the detailed knowledge of TL-efficiency to permit dose equivalent evaluation. The paper summarizes the theoretical models currently employed to describe TL-efficiency and reviews TL-efficiency measurements from the late 1960s up to now. Difficulties in comparing TL-efficiency data from various laboratories due to different readout systems and experimental protocols will be addressed. Valuable input for a critical discussion is provided by the large pool of recent TL-efficiency data generated in the framework of the ICCHIBAN intercomparison experiment. Particularly the results obtained by the authors for 6LiF:Mg, Ti (TLD-600), 7LiF:Mg, Ti (TLD-700), 6LiF:Mg, Cu, P (TLD-600H), 7LiF:Mg, Cu, P (TLD-700H) and CaF 2 : Tm (TLD-300) detectors, purchased from Thermo Fisher Scientific, Inc. (former Harshaw Chemical Co.), will be analyzed since these materials—and specific peaks in their glow curves—show distinctively different TL responses to HCP irradiation.

Radiation surveys and dose equivalent assessments for 60Co-contaminated rebar buildings

The first 60Co-contaminated rebar building was found in Taipei City in 1992. In order to provide effective remedial suggestions to the building owner and residents of contaminated housing units, and to evaluate the dose equivalent to the residents, the Atomic Energy Council (AEC) organized a survey team on October 17, 1992. The main object of this team was to perform a detailed survey and dose equivalent evaluation of the 60Co-contaminated buildings. By the end of 2000, 181 buildings with 1607 housing units were identified as being contaminated with radioactive rebars at different levels. This article presents the detailed radiation survey and dose equivalent evaluation procedures employed by the survey team.

Measurements of thorium–B ( [formula omitted]) in the outdoor environment and evaluation of equivalent dose

The activity size distribution of unattached as well as attached 212 Pb to aerosol particles was measured in the open air of El-Minia City, Egypt. The samples were collected using a wire screen diffusion battery technique and a low pressure cascade impactor. The mean activity median thermodynamic diameter (AMTD) of unattached 212 Pb was determined to be 1.4 nm with a relative mean geometric standard deviation ( σ g) of 1.55. A mean unattached fraction ( f p ) of 0.03±0.007 was obtained at a mean aerosol particle concentration of 32×10 3 cm −3. Sometimes the f p values were less than the detection limit of 0.008. The mean concentration of activity of 212 Pb was found to be 9.6±1.1 mBq m −3. The mean activity median aerodynamic diameter (AMAD) of the accumulation mode of attached 212 Pb was determined to be 360 nm with a mean ( σ g) of 2.7. The mean value of specific air activity concentration of 212 Pb associated with that mode was determined to be 303±12 mBq m −3. With a dosimetric model calculation (International Commission on Radiological Protection. Human respiratory tract model for radiological protection. Oxford: Pergamon Press, ICRP Publication 66, 1994) the total deposition fractions as well as total equivalent and effective dose have been evaluated considering the obtained parameters of the activity size distributions. At a total deposition fraction of about 97% for unattached activities the total equivalent and effective doses to the lung were determined to be about 0.18 and 0.02 μSv, respectively, while total equivalent and effective doses of about 0.45 and 0.05 μ Sv, respectively, were determined at a total deposition fraction of about 23% for the attached activities.

Rapid In-Situ Effective Dose Equivalent Measurement Using the Superheated-Liquid-Droplet Technique

This study performs a rapid in-situ effective dose equivalent evaluation using SLD. The effective dose equivalent recommended by ICRP-26 allows the direct calculation of the primary personnel limiting quantity by measuring the dose equivalent in various internal organs or tissues. Therefore, an indigenous multislab phantom coupled with the SLD multicounting system is set up to evaluate the effective dose equivalent, HE. Moreover, MCNP computational results are also compared. Comparisons indicate that underestimation of practical evaluations (within −15%) may be largely attributed to the insensitive nature of SLD in measuring the γ-ray induced dose.

Preliminary study of dose equivalent evaluation for residents in radioactivity contaminated rebar buildings

It has recently been found that several resident and office buildings in Taiwan were constructed with 60Co-contaminated reinforcing steel bar (rebar). Both governmental officials and the residents of such buildings have been concerned about this finding. In order to respond to the situation, the government has adopted a number of remedial measures, including full-scale radiation survey, dose evaluation and physical examinations of residents. This article presents three methods for evaluating the dose equivalents of the residents living in the contaminated rebar buildings by means of γ-ray survey, necklace-type thermoluminescence dosimeters (TLDs) and the human lymphocyte chromosome aberration analyses. The results reveal that the dose evaluation by γ-ray survey is rather conservative. Generally for the residents whose annual dose equivalents are greater than 5 mSv (0.5 rem) by γ-ray survey, the dose equivalents from necklace-type TLDs are only within the range of 20 to 50% of the evaluated values mentioned above. For chromosome analyses, at least 500 lymphocyte cells were scored and analyzed for each resident. Most of the chromosome analysis data show that the dose equivalents received by residents are lower than the detection limit of the method (100 mSv) and quite different from the estimated dose obtained from either γ-ray survey or necklace-type TLD measurements.

<sup>137</sup>Cs and <sup>134</sup>Cs Human Internal Contamination in Italy Following the 1986 Chernobyl Event

A synthesis of the data concerning the distribution and behaviour over time until September 1989 of 134Cs and 137Cs human contamination deriving from the accident to Unit 4 of the Chernobyl Nuclear Power Station on 26 April 1986, is presented. The controls have been performed by means of two intercalibrated whole-body counters on healthy adult male and female volunteers living in different Italian regions. The main topics investigated are: the behaviour over time of the contamination in Bologna and Rome: geographic distribution in Italy in September 1987; the statistical distribution of data; the variability of the individual activity in relation to the mean activity of homogeneous groups; the intersex differences; and the effect of the element's biokinetics uncertainties on the committed effective dose equivalent evaluation. Trends are also found, mainly useful where extended contamination involves too large a fraction of the population for individual control of all the subjects concerned, thus requiring the identification of small groups of individuals representative of large population groups.

A method to improve the evaluation of a combination track etch dosimeter/spectrometer

Approximate neutron energy spectra can be obtained by using a combination of dosimeter elements which have a different response for different neutrons energies. This paper describes a simple mathematical technique which can be applied to combination neutron dosimeter/spectrometers to increase the accuracy of the dose equivalent evaluation. A simple iterative algorithm is applied to determine the dose equivalent from CR-39 (polydiallyl-diglycol carbonate), polycarbonate, LR-115 (cellulose nitrate) track etch and albedo neutron detectors. The algorithm is simple enough to allow rapid evaluation of several different types of personnel neutron dosimeters with better accuracy than most existing methods.