RADIATION DOSE ASSESSMENT TO FAMILY MEMBERS TAKING CARE OF NON-CANCEROUS THYROID PATIENTS TREATED WITH I-131 THERAPY IN NUCLEAR MEDICINE DEPARTMENT

Abstract This study aimed to determine the effective doses of caregivers taking care of non-cancerous patients treated with iodine-131 (I-131). Patients (administered 185–1110 MBq of I-131) were given specific radiation safety instructions (RSI). Afterwards, caregivers were provided with thermoluminescent (TLD) dosimeter badges to be worn for 12–28 days when taking care of the patients. At the end of this period, TLD measurements were obtained. Results showed that caregivers’ mean effective dose was 0.15 ± 0.15 mSv, which is far less than the international recommendations of 5 mSv. Furthermore, the effective doses had no significant correlation with administered I-131 activity to the patients, distance from the hospital, caregivers’ age, educational level and mode of transport. Our study showed that radiation doses received by caregivers of non-cancerous patients are higher than that of cancerous patients, nevertheless their received doses were within the international limits, thereby indicating good compliance by the caregivers to RSI.

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PRELIMINARY EVALUATION OF PATIENT RADIATION DOSES DURING RADIONUCLIDE DIAGNOSTIC WITH MONOCLONAL ANTIBODIES LABELED WITH 89Zr

Pharmacy Formulas ◽

10.17816/phf81118 ◽

2021 ◽

Author(s):

Larisa A. Chipiga ◽

Anna E. Petrova ◽

Artem A. Mosunov ◽

Laura T. Naurzbaeva ◽

Stanislaus M. Kushnarenko ◽

...

Keyword(s):

Monoclonal Antibodies ◽

Dose Assessment ◽

Published Data ◽

Radiation Doses ◽

Preliminary Evaluation ◽

Software Products ◽

Absorbed Doses ◽

Order Of Magnitude ◽

Effective Doses

In connection with the constantly increasing use of monoclonal antibodies labeled with 89Zr, in clinical practice, it is urgent to study their pharmacokinetics with the determination, based on the data obtained, of absorbed doses in tumor foci, as well as intact organs and tissues, and effective doses of patients. To date, there are a limited number of studies that provide patient doses for diagnostic examinations using 89Zr-labeled monoclonal antibodies. In this regard, the purpose of this work was to assess the biodistribution of various monoclonal antibodies (ramucirumab, trastuzumab, atezolizumab) labeled with 89Zr, based on published data, with subsequent calculation of absorbed doses in radiosensitive organs and tissues and effective doses of patients. Based on the analysis of experimental data on the biodistribution of monoclonal antibodies labeled with 89Zr for the diagnosis of oncological diseases from the available literature sources and our own assessments, it has been concluded that the results of the determination of absorbed in organs and tissues and effective doses are inconsistent. The absorbed doses in organs, according to different literature sources, vary up to an order of magnitude within one organ and reach 440 mGy per examination, the effective dose varies from 3 to 112 mSv per examination. This may be due to differences in study design, radiometry and dose assessment methods. Comparison with doses obtained on the basis of a general model of biodistribution of monoclonal antibodies demonstrates the possibility of using this model for a rough estimate of internal doses of patients. However, for a more accurate assessment, it is necessary to standardize approaches to the determination of internal radiation doses using the most effective methodological solutions and software products.

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Modeling of the Radiation Doses during Dismantling of RBMK-1500 Reactor Pressurized Tanks from Emergency Core Cooling System

Science and Technology of Nuclear Installations ◽

10.1155/2013/576432 ◽

2013 ◽

Vol 2013 ◽

pp. 1-8 ◽

Cited By ~ 3

Author(s):

A. Simonis ◽

P. Poskas ◽

A. Sirvydas ◽

D. Grigaliuniene

Keyword(s):

Nuclear Power ◽

Radiation Safety ◽

Cooling System ◽

Computer Code ◽

Radiation Doses ◽

Planning Tool ◽

Direct Radiation ◽

Radioactive Aerosols ◽

Effective Doses ◽

Core Cooling

Decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. One of them is personnel radiation safety during the performance of dismantling activities. In this paper, modeling results of radiation doses during the dismantling of the pressurized tank from the emergency core cooling system (ECCS PT) of RBMK-1500 reactor are presented. The radiological surveys indicate that the inner surface of the ECCS PT is contaminated with radioactive products of corrosion and sediments due to the radioactive water. The effective doses to the workers have been modeled for different strategies of ECCS PT dismantling. In order to select the optimal personnel radiation safety, the modeling has been performed by the means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium). The impacts of dismantling tools, shielding types, and extract ventilation flow rate on effective doses during the dismantling of ECCS PT have been analyzed. The total effective personnel doses have been obtained by summarizing the effective personnel doses from various sources of exposure, that is, direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises. The uncertainty of the collective doses is also presented in this paper.

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Prediction of Radiation Doses During the Dismantling of the Pressurized Tank From Emergency Core Cooling System of RBMK-1500 Reactor

Volume 2: Facility Decontamination and Decommissioning; Environmental Remediation; Environmental Management/Public Involvement/Crosscutting Issues/Global Partnering ◽

10.1115/icem2013-96220 ◽

2013 ◽

Cited By ~ 2

Author(s):

A. Simonis ◽

P. Poskas ◽

G. Poskas

Keyword(s):

Nuclear Power ◽

Radiation Safety ◽

Cooling System ◽

Computer Code ◽

Radiation Doses ◽

Direct Radiation ◽

Radioactive Aerosols ◽

Effective Doses ◽

Core Cooling ◽

The Impact

Preparation for the decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. Personnel radiation safety during the performance of dismantling activities is one of them. In order to assess the optimal personnel radiation safety, the modelling is performed for large components by the means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium). Modelling results of radiation doses during the dismantling of the pressurized tank from the emergency core cooling system (ECCS PT) of RBMK-1500 reactor are presented in this paper. The mass of one ECCS PT is approximately 47.6 tons. The radiological surveys indicate that the inner surface of the ECCS PT is contaminated with radioactive products of corrosion and sediments due to the radioactive water. The assessment of workers exposure was performed to comply with ALARA. The effective doses to the workers were modeled for different strategies of ECCS PT dismantling. The impact of dismantling tools and shielding types and extract ventilation flow rate during the dismantling of ECCS PT on effective doses were analyzed. The total effective personnel doses were obtained by summarizing the effective personnel doses from various sources of exposure, i. e., direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises.

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Internal dose assessment of 148Gd using isotope ratios of gamma-emitting 146Gd or 153Gd in accidently released spallation target particles

Scientific Reports ◽

10.1038/s41598-020-77718-3 ◽

2020 ◽

Vol 10 (1) ◽

Author(s):

C. Rääf ◽

V. Barkauskas ◽

K. Eriksson Stenström ◽

C. Bernhardsson ◽

H. B. L. Pettersson

Keyword(s):

Operation Time ◽

Isotope Ratios ◽

Dose Assessment ◽

Sensitive Indicator ◽

Whole Body ◽

Internal Dose ◽

Tungsten Target ◽

Alpha Emitter ◽

Spallation Source ◽

Effective Doses

AbstractThe pure alpha emitter 148Gd may have a significant radiological impact in terms of internal dose to exposed humans in case of accidental releases from a spallation source using a tungsten target, such as the one to be used in the European Spallation Source (ESS). In this work we aim to present an approach to indirectly estimate the whole-body burden of 148Gd and the associated committed effective dose in exposed humans, by means of high-resolution gamma spectrometry of the gamma-emitting radiogadolinium isotopes 146Gd and 153Gd that are accompanied by 148Gd generated from the operation of the tungsten target. Theoretical minimum detectable whole-body activity (MDA) and associated internal doses from 148Gd are calculated using a combination of existing biokinetic models and recent computer simulation studies on the generated isotope ratios of 146Gd/148Gd and 153Gd/148Gd in the ESS target. Of the two gamma-emitting gadolinium isotopes, 146Gd is initially the most sensitive indicator of the presence of 148Gd if whole-body counting is performed within a month after the release, using the twin photo peaks of 146Gd centered at 115.4 keV (MDA < 1 Bq for ingested 148Gd, and < 25 Bq for inhaled 148Gd). The corresponding minimum detectable committed effective doses will be less than 1 µSv for ingested 148Gd, but substantially higher for inhaled 148Gd (up to 0.3 mSv), depending on operation time of the target prior to the release. However, a few months after an atmospheric release, 153Gd becomes a much more sensitive indicator of body burdens of 148Gd, with a minimum detectable committed effective doses ranging from 18 to 77 µSv for chronic ingestion and between 0.65 to 2.7 mSv for acute inhalation in connection to the release. The main issue with this indirect method for 148Gd internal dose estimation, is whether the primary photon peaks from 146 and 153Gd can be detected undisturbed. Preliminary simulations show that nuclides such as 182Ta may potentially create perturbations that could impair this evaluation method, and which impact needs to be further studied in future safety assessments of accidental target releases.

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MEASUREMENT OF THE ACTIVITY CONCENTRATIONS OF GAMMA EMITTING RADIONUCLIDES IN TOOTHPASTE SAMPLES AND ASSESSMENT OF THE CORRESPONDING ANNUAL EFFECTIVE DOSES

Radiation Protection Dosimetry ◽

10.1093/rpd/ncab038 ◽

2021 ◽

Author(s):

Anas M Ababneh ◽

Qutad M Samarah

Keyword(s):

Effective Dose ◽

Oral Hygiene ◽

Annual Effective Dose ◽

Daily Basis ◽

Dose Assessment ◽

Accidental Ingestion ◽

Activity Concentrations ◽

Effective Doses

Abstract It is inevitable that we are exposed to radiation daily from various sources and products that we consume on daily basis. The use of toothpaste for oral hygiene is one of the most common daily practices by humans and yet very little data are available regarding its radiation content. In this work, we investigated the concentrations of gamma emitting radionuclides in toothpaste samples consumed in Jordan. 40K and 226Ra were detected in almost one-third of the samples, whereas 228Ra was detected in nearly half of them. The corresponding activity concentrations in the detected samples were in the ranges of 68.7–154.2, 4.6–14.1 and 1.3–10.0 Bq/kg, respectively. Dose assessment of accidental ingestion of toothpaste for children and adults was made, and its contribution to the annual effective dose was found to be very minimal with maximum doses of ~2.9 and 1.3 μSv for children and adults, respectively.

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