ICRP Recommendations Research Articles

Assessment of the skin contamination dose coefficients for 252Cf radionuclide: Monte Carlo approach

Handling the 252Cf radionuclide source poses a potential hazard of skin surface contamination in case of an unexpected occurrence. Consequently, there is a growing need to establish precise dose conversion coefficients tailored to each type of emitted primary particle and various radionuclides. Nevertheless, the current body of literature does not provide specific data or methodologies for evaluating skin contamination dose and its associated coefficients, particularly with regard to the 252Cf source. Thus, this study aims to quantify the dose rate received by the skin and its associated coefficients after contamination scenario. Utilizing the established MCNPX environment, the Equivalent dose rate and Absorbed dose, along with Skin contamination dose coefficient (SCDC), have been calculated within the skin tissue. Two methodologies, specifically Watt Fission distribution and the Doppler Effect, are proposed to analyze particle spectra within skin phantom, enabling the calculation of Equivalent dose rate. In accordance with ICRP recommendations regarding the optimal depth for assessing skin doses, the designated scoring volume within the skin is located between depths of 50–100 μm. This volume is tasked with evaluating the dose. The SCDC results were entirely consistent with previously published data from MCNPX, with statistical uncertainties of less than 15%, demonstrating the efficacy of the methodologies employed in this study. This research presents an innovative method for generating data related to skin contamination doses. The novel outcomes in the current research facilitate the assessment of skin dose contamination for the targeted radionuclides and radiotherapy purposes due to staff oversight and radiobiological effects.

Radiation dose assessments for generic nuclear power plants part I: Routine operation

The main task of the site evaluation report for nuclear installation concerning nuclear safety is the protection of the public and environment from the radiological consequences of radioactive releases in both cases of normal operation and accident conditions. The atmosphere is the most important pathway to be considered in the assessment of the environmental impact of radioactive materials released from nuclear facilities. The present study is a comprehensive investigation of environmental assessments for the dose calculation model resulting from routine operation NPPs of 1000 Mw (e). The procedure consists of different parts, beginning with the manipulation of collected meteorological data such as wind speed and direction, temperature, incoming solar radiation and utilizing a set of empirical formulae for evaluating night solar radiation emitted from the ground, and then evaluating hourly stability classes and joint frequency distribution of winds by developing code. The dilution factor was evaluated using computer code (XOQDOQ code). Finally, the radiation dose assessments resulting from the routine operation of NPPs were calculated. The processes are adapted with IAEA recommendations documents, safety guides, and ICRP recommendations. The results reveal that there is no detectable value that affects the people surrounding the site nor the environmental area concerning low population zone or exclusion area.

Analysis of natural occurring radionuclide activity concentration of fish in west coast waters of Sabah, Malaysia

Abstract Fish is an important source of protein in human diets, but concerns arise due to natural radionuclide contamination in food and water sources. The study aimed to assess the concentration radionuclides activities (232Th, 238U, and 40K) in three commercial fish species from west coast waters of Sabah, Malaysia. Additionally, the annual effective dose and cancer risk for adults consuming these fish were evaluated. The concentration analysis was conducted using the inductively coupled plasma-mass spectrometer (ICP-MS) technique. The average concentration levels of radionuclide activity detected, and the annual effective dose was estimated to be much lower than UNSCEAR recommendations. The collective effective dose was estimated at 1.88 μSv y−1 for 232Th, 1.11 μSv y−1 for 238U, and 15.12 μSv y−1 for 40K. The cancer risk for adults from the annual effective dose was also found to be much lower than UNSCEAR and ICRP recommendations. Based on the study’s findings, consuming fish from west coast waters of Sabah is deemed safe. However, it is recommended to establish a long-term monitoring system for radionuclide bioaccumulation in fish to gather valuable information for assessing the potential health risks associated with radionuclides in Malaysia, particularly in the Sabah.

Measurement of Background Ionizing Radiation Around Beta Glass Plc and Its Environs, Ughelli, Delta State, Nigeria

A study of the background ionizing radiation levels to helps monitor the impact of human activities on the environment, such as mining, farming, glass bottle production and power generation has been carried out around Beta glass Plc and Transcorp Power LTD using a hand-held gamma spectrometer (GAMMA-SCOUT; Geiger counter Radiation meter) and GPS (Oregon-450 Garmin). The study of the BIR levels was carried out between June to July, 2022. The average measured exposure rates (AV. BIR) ranged from 0.007 mRh-1 to 0.013 mRh-1, with 0.010±0.0002 mRh-1 as the average. The estimated mean outdoor absorbed dose rate (ADR) for the study region was 113.10 Gyh-1 to 87.153±2.091 Gyh-1, with an average annual effective dose equivalent (AEDE) of 106.884±2.565 mSvy-1 and an average excess lifetime cancer risk (ELCR) of (0.374±0.009) × 10-3. The measured values of radiological parameters, such as annual dose rate, annual effective dose rate, and excess lifetime cancer, were higher than background radiation levels. The average BIR is in line with the ICRP recommendation of 0.013 (mRh-1), while the effective dose rate (EDR) is somewhat below the ICRP recommendation of 1.00 (mSvy-1) for a normal environment. The obtained values of the radiological indicators, such as ADR, AEDE, and ELCR, are greater than the background radiation norm of 59 (?Gyh-1), 0. 07 (mSvy-1), and 0.29×10-3. The study found that radiation levels and doses in the study location do not immediately harm personnel or residents, but increased lifetime cancer risk values suggest that cumulative dosages may cause cancer.

Assessment of the Natural Radioactivity Levels in Soils and the Associated Health Risks to the Public in Mbeya City, Tanzania

The objective of this study was to assess the radioactivity levels in soils from Mbeya City and their associated health risks. Samples were collected from selected locations. Then, the samples were analyzed by using an HPGe detector to determine the concentrations of 232Th, 226Ra, and 40K. The concentrations were used to compute the radiation indices. The concentrations ranged from 13.98±1.99 to 275.93±25.88 Bq/kg for 232Th, 5.69±1.48 to 107.49±10.61 Bq/kg for 226Ra and 195.76±21.17 to 1,710.00±161.32 Bq/kg for 40K with an average of 169.55 Bq/kg, 71.86 Bq/kg and 998.42 Bq/kg, respectively. The radium equivalent ranged from 41.10 to 583.93 Bq/kg with an average of 391.19 Bq/kg. The gamma dose rate ranged from 42.97 to 691.85 nGyhr-1 with an average of 467.57 nGyhr-1. The indoor effective dose ranged from 0.21 to 3.39 mSv/y, with an average of 2.29 mSv/y while the outdoor effective dose ranged from 0.05 to 0.85 mSv/y, with an average of 0.57 mSv/y. The internal hazard index ranged from 0.13 to 1.84 Bq/kg with an average of 1.25 Bq/kg while the external hazard index ranged from 0.11 to 1.58 Bq/kg with an average of 1.06.Bq/kg. The results indicate elevated levels of background radiation above the world average and the hazard indices above the ICRP and UNSCEAR recommendations. Therefore, control measures were recommended to protect the public against possible health risks.

Estimation of Radon/Thoron and their progeny concentration in some dwellings and exhalation rates from soils of Samana city, Patiala

The study of indoor pollutants and radioactive nuclides imposes importance on human health as we spend most of our time indoors. Out of all the natural radioactive elements, Radon-222 and Radon220 are present in large abundance indoors. Both are radioactive and decay to various short-lived progenies by emitting high-energy alpha particles, so inhalation of such gases can harm the human body in various ways. In the present report, the concentration of Rn-222 and Rn-220 and its progenies are estimated from some dwellings of Samana city of Patiala district, Punjab. Pin-hole dosimeters with DRPS and DTPS were employed in 15 houses in the study area to measure indoor Rn-222 and Rn-220 concentrations. The concentration of Rn-222 ranges between 401.96 Bq/m3 and 27.36 Bq/m3 with mean value of 138.82 Bq/m3, and the concentration of Rn-220 varies from 222.2 Bq/ m3 to 50 Bq/m3 with the mean weight of 131.12 Bq/m3. The average inhalation dose was also calculated from the results and was found to have an average value of 2.19 mSv per year. The average weight of radon concentration was below the ICRP recommendation of 200-300 Bq/m3 but the value is above these limits at some locations. Radon and thoron emission rates from soil samples near to the dwellings was also calculated.

The energy and angular Hp(3,α) response of commercial eye-lens dosemeters to photon and beta particle radiation

In 2012, the International Commission on Radiological Protection recommended a reduction of the equivalent dose limit for the lens of the eye from 150 mSv to 50 mSv per year, with a maximum 5-year average of 100 mSv. This noticeable reduction imposes more demands on the performance on personnel eye-lens dosimetry. Regulatory authorities of various jurisdictions worldwide followed some or all, of the ICRP recommendations and implemented reduced occupational lens of eye dose limits in their legislation. Commercial dosimetry service providers have risen to the challenges by developing new dosemeter technologies, and some in-house dosimetry services have developed their own eye-lens dosemeters. The results of the energy and angular response, to photon and beta radiation, of three commercial dosemeters are reported here. Their performances are compared to those of the dosemeter developed, in-house, by Canadian Nuclear Laboratories. With some adjustments to their calibrations, the responses of all dosemeters, could be within the performance limits set by the ISO 14146 and the IEC 62387 standards for the conditions of the tests that were performed.

Assessment of reference dose levels for nuclear medicine center in Jordan

This study aims to evaluate diagnostic reference levels (DRLs) for nuclear medicine (NM) imaging studies in Jordan. The parameters of the current study obtained from different NM modalities conducted for adult patients imaging. The Administrated Activity (AA) of a radioactive substance (MBq) was collected from four hospitals specialized in nuclear medicine imaging at Amman City. According to ICRP recommendation, the established DRLs values obtained based on the 75th percentile of the AA values. The obtained results were compared with several international results. The DRLs' values for Jordan were located within the recommended range.

Particle Size Dependent Dissolution of Uranium Aerosols in Simulated Gastrointestinal Fluids.

Uranium aerosol exposure can be a health risk factor for workers in the nuclear fuel industry. Good knowledge about aerosol dissolution and absorption characteristics in the gastrointestinal tract is imperative for solid dose assessments and risk management. In this study, an in vitro dissolution model of the GI tract was used to experimentally study solubility of size-fractionated aerosols. The aerosols were collected from four major workshops in a nuclear fuel fabrication plant where uranium compounds such as uranium hexafluoride (UF 6 ), uranium dioxide (UO 2 ), ammonium uranyl carbonate, AUC [UO 2 CO 3 ·2(NH 4 ) 2 CO 3 ] and triuranium octoxide (U 3 O 8 ) are present. The alimentary tract transfer factor, f A , was estimated for the aerosols sampled in the study. The transfer factor was derived from the dissolution in the small intestine in combination with data on absorption of soluble uranium. Results from the conversion workshop indicated a f A in line with what is recommended (0.004) by the ICRP for inhalation exposure to Type M materials. Obtained transfer factors, f A , for the powder preparation and pelletizing workshops where UO 2 and U 3 O 8 are handled are lower for inhalation and much lower for ingestion than those recommended by the ICRP for Type M/S materials f A = 0.00029 and 0.00016 vs. 0.0006 and 0.002, respectively. The results for ingestion and inhalation f A indicate that ICRP's conservative recommendation of f A for inhalation exposure is applicable to both ingestion and inhalation of insoluble material in this study. The dissolution- and subsequent absorption-dependence on particle size showed correlation only for one of the workshops (pelletizing). The absence of correlation at the other workshops may be an effect of multiple chemical compounds with different size distribution and/or the reported presence of agglomerated particles at higher cut points having more impact on the dissolution than particle size. The impact on dose coefficients [committed effective dose (CED) per Bq] of using experimental f A vs. using default f A recommended by the ICRP for the uranium compounds of interest for inhalation exposure was not significant for any of the workshops. However, a significant impact on CED for ingestion exposure was observed for all workshops when comparing with CED estimated for insoluble material using ICRP default f A . This indicates that the use of experimentally derived site-specific f A can improve dose assessments. It is essential to acquire site-specific estimates of the dissolution and absorption of uranium aerosols as this provides more realistic and accurate dose- and risk-estimates of worker exposure. In this study, the results indicate that ICRP's recommendations for ingestion of insoluble material might overestimate absorption and that the lower f A found for inhalation could be more realistic for both inhalation and ingestion of insoluble material.

ОДЭК: уровень радиологической защиты населения с учётом современной модели «доза-эффект» МКРЗ

After receiving the notification of the license granting and its registration in the license register it has become topical the preparation of the final version of the project documentation to guarantee the safety of the public residing in the proximity of the operating Pilot-demonstration energy complex (PDEC), which includes a reactor unit BREST-OD-300, reprocessing module and a fuel fabrication/refabrication module. Currently International Commission on Radiological Protection (ICRP, Publication 103) recommends for estimating radiological protection of the public to use carcinogenic risks estimates for individual organs and tissues, with account of equivalent doses and modern models of «dose-effect» relationship instead of the use of committed effective doses. The article demonstrates the 5-7 times spread of lifetime carcinogenic radiation risk values from exposure to different radionuclides, while the committed effective dose from intakes of the same radionuclides do not differ. The distinguishing characteristic of updated ICRP recommendations is consideration of gender and age of the exposed population when assessing radiation-related carcinogenic risks. The article presents estimates of lifetime attributable risk for critical groups of the public residing in the proximity of the PDEC; the risk was calculated with the use of estimates of annual air emissions from normally operated PDEC. Tritium (H-3) and cesium (Cs-137) emissions limit values have been calculated, the radionuclides ensure the normal functioning of the complex in a negligible radiation risk range for the population (Radiation Safety Standards (RSS-99/2009)).

ПРИМЕНЕНИЕ НОВЫХ РЕКОМЕНДАЦИЙ МКРЗ ДЛЯ РАСЧЕТА ДОЗ ПЕРСОНАЛА ПРИ ИНГАЛЯЦИОННОМ ПОСТУПЛЕНИИ РАДИОНУКЛИДОВ УРАНА

Purpose: Harmonization and improvement of the system for regulating the internal radiation exposure of workers and the basic requirements for ensuring radiation safety, taking into account the application of new international requirements and recommendations. Material and methods: A brief description of the procedure for calculating absorbed and equivalent doses in organs and tissues after the intake of radionuclides into the human body is presented, using the biokinetic and dosimetric models adopted in the new ICRP recommendations, as well as a discussion of the impact of these changes on the results of calculating dose coefficients for the case of inhalation intake of uranium-235 radionuclide. Results: The effective dose values and equivalent doses to organs and tissues for workers were calculated depending on the time after a single inhalation intake of uranium-235 aerosol, according to new models [1–12] and according to previous ICRP models [15, 16]. The calculation of the effective dose according to the new models included calculations of equivalent doses for 14 main organs and tissues and 13 organs and tissues classified as “remainder tissues” as described in ICRP Publication 103 [3]. The committed effective dose was then calculated according to the new approach using the average of the equivalent doses for the reference adult male, HTM, and the reference adult female, HTF, as well as the tissue and organ weighting factors, WT, adopted in ICRP Publication 103. The values of the effective dose and equivalent doses on the red bone marrow, lungs and remainder tissues vs time in the range from several days to 18250 days (50 years) after a single inhalation intake of an aerosol of uranium-235 for standard value AMAD=5 µm and types of compounds F, M, S, F/M and M/S are presented according to new and previous ICRP models. It is shown that the value of the dose coefficient for type F, calculated by new models, is 2.6 times (2.3E-07÷6.0E-07) less than that calculated by previous ICRP models, and the value of the dose coefficient for type F/M calculated by new models is 1.6 times (3.8E-07÷6.0E-07) less than the value of the dose coefficient for type F calculated by previous ICRP models. For uranium trioxide UO3, taking into account its transition from compound type M to F/M, the value of the dose coefficient for committed effective dose according to the updated model of the respiratory tract is 4.7 times (3.8E-07÷1.8E-06) less than the corresponding value for the previous model of the type M respiratory tract. The committed effective dose value for compound type M, calculated using the new models, is 1.4 times (1.3E-06÷1.8E-06) less than the same value calculated using the previous ICRP models. The value of the committed effective dose for type M/S compounds (which, according to the new model of the respiratory tract, include uranium oxide U3O8 and dioxide UO2), calculated according to new models, is 1.2 times (5.1E-06÷6.1E-06) less than the value calculated from previous ICRP models for type S compounds (which included U3O8 and UO2 in the previous respiratory tract model). Conclusion: From the above data it follows that in case of the adoption of national radiation safety standards to new ICRP models, differences in the values of dose coefficients will result in a change of annual limits of intake (ALI) in the corresponding proportion for the types of uranium aerosol compounds noted above.

CT scanner-specific organ dose coefficients generated by Monte Carlo calculation for the ICRP adult male and female reference computational phantoms

Objective. Provide analyses of new organ dose coefficients (hereafter also referred to as normalized doses) for CT that have been developed to update the widely-utilized collection of data published 30 years ago in NRPB-SR250. Approach. In order to reflect changes in technology, and also ICRP recommendations concerning use of the computational phantoms adult male (AM) and adult female (AF), 102 series of new Monte Carlo simulations have been performed covering the range of operating conditions for 12 contemporary models of CT scanner from 4 manufacturers. Normalized doses (relative to free air on axis) have been determined for 39 organs, and for every 8 mm or 4.84 mm slab of AM and AF, respectively. Main results. Analyses of results confirm the significant influence (by up to a few tens of percent), on values of normalized organ (or contributions to effective dose (E 103,phan)), for whole body exposure arising from selection of tube voltage and beam shaping filter. Use of partial (when available) rather than a Full fan beam reduced both organ and effective dose by up to 7%. Normalized doses to AF were larger than corresponding figures for AM by up to 30% for organs and by 10% for E 103,phan. Additional simulations for whole body exposure have also demonstrated that: practical simplifications in the main modelling (point source, single slice thickness, neglect of patient couch and immobility of phantom arms) have sufficiently small (<5%) effect on E 103,phan; mis-centring of the phantom away from the axis of rotation by 5 mm (in any direction) leads to changes in normalized organ dose and E 103,phan by up to 20% and 6%, respectively; and angular tube current modulation can result in reductions by up to 35% and <15% in normalized organ dose and E 103,phan, respectively, for 100% cosine variation. Significance. These analyses help advance understanding of the influence of operational scanner settings on organ dose coefficients for contemporary CT, in support of improved patient protection. The results will allow the future development of a new dose estimation tool.

Analysis of correspondence between the current individual monitoring system of internal exposure caused by plutonium intake and the relevant ICRP recommendations

Individual dosimetry monitoring of occupational internal exposure due to plutonium intake is performed by the bioassay method which measures the plutonium isotopes in daily urine samples. Then, to estimate normalized values (committed effective dose and intake), biokinetic and dosimetry models are used. The currently valid major instruments of radiation safety in the Russian Federation are based on the requirements and approaches outlined in Publications 68 and 78 of the International Commission on Radiobiological Protection. In implementing the requirements of Publication 130 there will be a need for updating the Russian guidance documents to the new requirements. The purpose of this study was the analysis of correspondence of the current system of individual monitoring of internal exposure caused by plutonium intake to the relevant recommendations of the International Commission on Radiobiological Protection outlined in Publication 130 by the example of the FSUE “Mayak PA” production conditions. As a result of the study, the values of annual limit of plutonium intake were determined based on new biokitetic and dosimetry models, and the measured values were calculated during the current dosimetry monitoring corresponding to the value of committed effective dose limit for Category A, B personnel and women under 45 years of age. It was found that following the introduction of new requirements of the International Commission on Radiobiological Protection for Type M compounds, an intake of more than 1/4 of annual limit of plutonium intake could be determined within a year after intake using the method of alpha spectrometry. To determine an intake of 1/20 of annual limit of intake at the recommended interval of 180 days between measurements, it is necessary to use inductively-coupled plasma mass spectrometry. For the personnel working with insoluble compounds of plutonium, an intake of one annual limit of intake and 1/4 of annual limit of intake can be determined using mass spectrometry, and 1/20 of annual limit of intake can be determined using only a more sensitive method.

Estimation of the total body burden for some individuals using whole body counter

The current research study aims to estimate the total body burden and internal contamination for some public and occupational workers at Nuclear Research Center, NRC, and Egypt using Accuscan II whole body counter. Material and Method WBC consist of three parts; 25% coaxial germanium detector connected with electronic circuits, Shadow shield and transfer phantom. The WBC was used to monitor the workers and public for the 99mTc, 131I and 40K in the total body geometry for acquisition time of 300 sec. The efficiency calibration was carried out for wide energy range from 59 keV to 1836.2 keV using a standard mixed source. A liquid vial mixed gamma source inserted in thyroid and total body geometry positions of transfer phantom. In addition, KCl salt is inserted in total body geometry position of transfer phantom in real geometry to confirm the calibration of 40K in the energy of 1460 keV. Results The minimum detectable amount for 99mTc, 131I and 40K was 178, 172 and 204 Bq respectively. 40K body content activity for males ranged from 3497 to 5594 Bq and specific activity of 40K was ranged from 43.7 to 75.3 Bq/Kg and for females ranged from 2797 to 4544 Bq and 40K specific activity ranged from 31.1 to 64 Bq/Kg. The annual internal effective dose from 40K for males ranged from 122 to 210 μSv with an average dose of 165 μSv and for females ranged from 86 to 163 μSv with an average dose of 135 μSv. Internal dose for individuals exposed to I131 ranged from 0.45 μSv to 168.5 μSv and for individuals exposed to 99mTc ranged from 1.78 μSv to 3.25 μSv. The estimated annual effective dose for 40K, 131I and 99mTc was within the ICRP recommendations.

Manual on the proper use of meta-[211At] astato-benzylguanidine ([211At] MABG) injections in clinical trials for targeted alpha therapy (1st edition)

In this manuscript, we present the guideline for use of meta-[211At] astatobenzylguanidine ([211At] MABG), a newly introduced alpha emitting radiopharmaceutical to the up-coming World’s first clinical trial for targeted alpha therapy (TAT) at Fukushima Medical University in Japan, focusing on radiation safety issues in Japan. This guideline was prepared based on a study supported by the Ministry of Health, Labour, and Welfare, and approved by the Japanese Society of Nuclear Medicine on Oct. 5th, 2021. The study showed that patients receiving [211At] MABG do not need to be admitted to a radiotherapy room and that TAT using [211At] MABG is possible on an outpatient basis. The radiation exposure from the patient is within the safety standards of the ICRP and IAEA recommendations for the general public and caregivers or patient’s family members. In this guideline, the following contents are also included: precautions for patients and their families, safety management associated with the use of [211At] MABG, education and training, and disposal of medical radioactive contaminants. TAT using [211At] MABG in Japan should be carried out according to this guideline. Although this guideline is based on the medical environment and laws and regulations in Japan, the issues for radiation protection and evaluation methodology presented in this guideline are useful and internationally acceptable as well.

A study on pressurizer cutting scenario for radiation dose reduction for workers using VISIPLAN

The operations in the design lifecycle of a nuclear power plant targeted to be decommissioned lead to neutron activation. Operations in the decommissioning process include cutting, decontamination, disposal, and processing. Among these, cutting is done close to the target material, and thus workers are exposed to radiation. As there are only a few studies on pressurizers, there arises the need for further research to assess the radiation exposure dose. This study obtained the specifications of the AP1000 pressurizer of Westinghouse and the distribution of radionuclide inventory of a pressurizer in a pressurised water reactor for evaluation based on literature studies. A cutting scenario was created to develop an optimal method so that the cut pieces fill a radioactive solid waste drum with dimensions 0.571 m × 0.834 m. The estimated exposure dose, estimated using the tool VISIPLAN SW, in terms of the decontamination factor (DF) ranged from DF-0 to DF-100, indicating that DF-90 and DF-100 meet the ICRP recommendation on exposure dose 0.0057 mSv/h. At the end of the study, although flame cutting was considered the most efficient method in terms of cutting speed, laser cutting was the most reasonable one in terms of the financial aspects and secondary waste.

ГАРМОНИЗАЦИЯ ЗАКОНОДАТЕЛЬНЫХ АКТОВ РОССИЙСКОЙ ФЕДЕРАЦИИ С СОВРЕМЕННЫМИ МЕЖДУНАРОДНЫМИ РЕКОМЕНДАЦИЯМИ

Purpose: Harmonization of the Russian Federation legislation with current international recommendations&#x0D; Results: The concept of the radiation safety system has been significantly modified by recommendations of ICRP (2007) and IAEA (2014). An analysis of existing international regulatory framework for radiation safety allowed to identify the main provisions to be implemented in the Russian legal and regulatory framework. It’s showed that the current Federal Law of 09.01.1996 No. 3-FZ «On Radiation Safety of Population» must be ultimately revised to be harmonized with international documents. General approaches to legal regulation of radiation safety should be essentially modified to create a strong relationship between this law and other regulatory and legal documents in force in the Russian Federation.&#x0D; Conclusion: An article-by-article analysis of the current Federal Law of 09.01.1996 No. 3-FZ «On Radiation Safety of Population « showed the need to modify 22 existing articles and add 12 new articles in order to harmonize it with international documents. Given such a large volume of modification it is advisable to pass a new law with simultaneous abolition of the current federal law. A new name has been proposed: Federal Law of the Russian Federation «On Radiation Safety in the Russian Federation». &#x0D; The enactment of the Federal Law of the Russian Federation «On Radiation Safety in the Russian Federation» with the main by-laws approved by the Russian Federation Government – «Radiation Safety Standards» and «Basic Rules for Ensuring Radiation Safety» – will allow to establish an actual regulatory framework for ensuring radiation safety of personnel and population in Russia.

СРАВНИТЕЛЬНЫЙ АНАЛИЗ ПОДХОДОВ К НОРМИРОВАНИЮ И КОНТРОЛЮ ВНУТРЕННЕГО ОБЛУЧЕНИЯ ПЕРСОНАЛА

Purpose: Harmonization and improvement of the system for regulating the internal radiation exposure of workers and the basic requirements for ensuring radiation safety with international requirements and recommendations. Material and methods: Issues related to the development of approaches to regulation and monitoring of workers for internal radiation exposure in the process of evolution of the ICRP recommendations and the national radiation safety standards, are considered. The subject of analysis is the standardized values: dose limits for workers and permissible levels as well as directly related methods of monitoring of workers for internal radiation exposure, whose purpose is to determine the degree of compliance with the principles of radiation safety and regulatory requirements, including non-exceeding the basic dose limits and permissible levels. The permissible levels of inhalation intake of insoluble compounds (dioxide) of plutonium-239 are considered as a numerical example. Results: Based on the analysis of approaches to the regulation and monitoring of workers for internal radiation exposure for the period from 1959 to 2019, it is shown that a qualitative change in the approach occurred in the 1990s. It was due to a decrease in the number of standardized values by introducing a single dose limit for all types of exposure: the effective dose E, which takes into account the different sensitivity of organs and tissues for stochastic radiation effects (WT), using the previously accepted concepts of the equivalent dose H and groups of critical organs. From the analysis it follows that the committed effective dose is a linear transformation of the intake, linking these two quantities by the dose coefficient, which does not depend on the time during which the intake occurred, and reflects certain exposure conditions of the radionuclide intake (intake routes, parameters of aerosols and type of radionuclide compounds). It was also shown that the reference value of the function z(t) linking the measured value of activity in an organ (tissue) or in excretion products with the committed effective dose for a reference person, which is introduced for the first time in the publications of the ICRP OIR 2015-2019, makes it possible to standardize the method of measuring the normalized value of the effective dose. Based on the comparison of the predicted values of the lung and daily urine excretion activities following constant chronic inhalation intake of insoluble plutonium compounds at a rate equal annual limit of intake (ALI) during the period of occupational activity 50 years it was shown that the modern biokinetic models give a slightly lower level (on average 2 times) of the lungs exposure compared to the models of the previous generation and a proportionally lower level (on average 1.4 times) of plutonium urine excretion for the standard type of insoluble plutonium compounds S. However, for the specially defined insoluble plutonium compound, PuO2, the level of plutonium urine excretion differs significantly downward (on average 11.5 times) compared to the models of the previous generation. Conclusion: With the practical implementation of new ICRP OIR models, in particular for PuO2 compounds, additional studies should be carried out on the behavior of insoluble industrial plutonium compounds in the human body. Besides, additional possibilities should be used to determine the intake of plutonium by measuring in the human body the radionuclide Am-241, which is the Pu-241 daughter. To determine the plutonium urine excretion, the most sensitive measurement techniques should be used, having a decision threshold about fractions of mBq in a daily urine for S-type compounds and an order of magnitude lower for PuO2 compounds. This may require the development and implementation in monitoring practice the plutonium-DTPA Biokinetic Model.

A Review of the Latest ICRP Recommendations on Effective Dose: Commentary of ICRP Publication 147

A Review of the Latest ICRP Recommendations on Effective Dose: Commentary of ICRP Publication 147

Establishment of typical adult CT dose indicators for PET-CT scans in Slovenia

The aim was to determine typical values of diagnostic reference level (DRL) quantities for the computed tomography (CT) part of the most common positron emission tomography—computed tomography (PET-CT) procedures in Slovenia. The most common PET-CT procedures were identified, and data collated for 565 patients imaged in all three PET-CT units in Slovenia during a time span of 11 months. As the number of facilities is too low to establish national DRLs, we followed ICRP recommendations and determined typical values of DRL quantities as the median values of the pooled set of data. Mean, median, and standard deviation of CT dose index (CTDIvol) and total dose length product (DLP) for the CT part of the most common PET-CT procedures were determined for pooled data as well as for each PET-CT unit. The data were compared between all three units to identify possible outliers that would likely benefit from optimization. Three most common CT protocols covering approximately 2/3 of all PET-CT imaging performed in Slovenia were considered: from the base of the cranium to the middle of the femur, from the top of the head to the middle of the femur, and for the whole-body PET-CT. The established typical values in terms of total DLP were 295, 359, and 676 mGyċcm, respectively; and in terms of CTDIvol 3.05, 3.22, and 3.60 mGy, respectively. Comparing the data between all three units showed significantly higher (p < 0.001) patient doses on one unit, indicating a need for optimization. The results present the first-time data on the national typical values of DRL quantities for the CT part of most common PET-CT procedures in Slovenia. While the determined typical values are within the DRL values established in some other countries, significant differences were found between the individual units included in the study.