International Commission On Radiological Protection Publication Research Articles

APPLICATION OF INTDOSKIT TOOL FOR ASSESSMENT OF UNCERTAINTIES ON DOSE COEFFICIENTS FOR INGESTION OF URANIUM BY WORKERS.

The International Commission on Radiological Protection (ICRP) has developed the Human Alimentary Tract Model (HATM) to calculate radiation doses from the ingestion of radionuclides for the protection of workers and the public. In parallel, the ICRP's Occupational Intake of Radionuclides (OIR) series provides biokinetic models and dose coefficients based on a reference human, primarily for regulatory purposes. Although these coefficients are not usually checked for uncertainties, the investigation of such uncertainties is crucial to ensure their reliability in radiation protection. This study uses INTDOSKIT, a software tool developed using the R programming language and RStudio as the Integrated Development Environment (IDE), to calculate doses and explore uncertainties following ingestion of U-238 by workers. Two scenarios were investigated: Case I, using the latest HATM model and the systemic uranium model from ICRP publications, and Case II, based on data from Puncher and Burt (2013), using older ICRP models. In both cases, uptake of U-238 in its soluble form (Type F) was modelled and the results were validated using the ICRP OIR dataviewer software. Following validation, uncertainty and sensitivity analyses were performed. In Case I, uncertainties were assigned to particle transport parameters in the HATM and uranium uptake fraction (fA) into the blood, while in Case II they were limited to the uptake fraction (f1) and the systemic uranium model. A Monte Carlo simulation of 60,000 runs was performed for both cases, sampling model parameters from their respective probability distributions to generate dose distributions. The influence of each parameter on these distributions was also analyzed. Probability distributions were inferred to the calculated dose values using the maximum likelihood estimation method and the Kolmogorov-Smirnov goodness-of-fit. The results showed that for both cases the committed effective dose coefficient, e (50), followed a lognormal distribution. Case I had a geometric mean of 3.2E-08 Sv/Bq and GSD of 2.0, while case II had a slightly lower geometric mean of 3.1E-08 Sv/Bq and GSD of 1.9. Sensitivity analysis showed that the main contributor to dose uncertainty was the fraction of uranium absorbed from the small intestine into the blood. Both cases showed similar trends, with slightly higher results in case I. Overall, this study demonstrates the effectiveness of INTDOSKIT in calculating dose coefficients and analyzing uncertainties. It suggests that while the ICRP reference values remain useful for protection, the incorporation of additional statistical measures and distribution characteristics could further enhance radiation protection strategies.

Individualized dose calculation for internal exposure on radionuclide intake: GPU acceleration approach

Objective. The rapid and accurate assessment of internal exposure dose is a crucial safeguard for personnel health and safety. This study aims to investigate a precise and efficient GPU Monte Carlo simulation approach for internal exposure dose calculation. It directly calculates doses from common radioactive nuclides intake, like 60Co for occupational exposure, allowing personalized assessments. Approach. This study developed a GPU-accelerated Monte Carlo program for internal exposure on radionuclide intake, successfully realizing photoelectronic coupled transport, nuclide simulation, and optimized acceleration. The generation of internal irradiation sources and sampling methods were achieved, along with the establishment of a personalized phantom construction process. Three irradiation scenarios were simulated to assess computational accuracy and efficiency, and to investigate the influence of posture variations on internal dose estimations. Main results. Using the International Commission on Radiological Protection (ICRP) voxel-type phantom, the internal dose of radionuclides in individual organs was calculated, exhibiting relative deviation of less than 3% in comparison to organ dose results interpolated by Specific Absorbed Fractions in ICRP Publication 133. Employing the Chinese reference phantom for calculating internal irradiation dose from the intake of various radionuclides, the use of GPU Monte Carlo program significantly shortened the simulation time compared to using CPU programs, by a factor of 150–500. Internal dose estimation utilizing a seated Chinese phantom revealed up to a 75% maximum difference in organ dose compared to the same phantom in a standing posture. Significance. This study presents a rapid GPU-based simulation method for internal irradiation doses, capable of directly simulating dose outcomes from nuclide intake and accommodating individualized phantoms for more realistic and expeditious calculations tailored to specific internal irradiation scenarios. It provides an effective and feasible tool for precisely calculating internal irradiation doses in real-world scenarios.

Improving the reliability of internal dosimetry for uranium workers: Application of the ICRP/OIR uranium model to long-term occupational intakes

In the Occupational Intakes of Radionuclides (OIR) report series, the International Commission on Radiological Protection (ICRP) published updated biokinetic models and dosimetric data associated with the internal exposure of workers, which are consistent with the recommendations of ICRP Publication 103. The present study focused on the application of the new OIR uranium model published in ICRP Publication 137, which is relevant for the interpretation of measurements of activity in the body and in excreta, and for calculating the committed effective dose. The main objective of this study was to assess the impacts of the new OIR biokinetic models and dose coefficients compared with the old ones based on ICRP Publications 60/78/119. CIEMAT and the University of Salamanca in Spain have been working on re-interpreting in vitro bioassay data obtained from workers with long-term exposure to inhaling uranium oxides during the fabrication of nuclear fuel elements using low enriched uranium, which is now classified as OIR type M/S material. The effects on the retention/excretion models and dose coefficients were studied for 234U, 235U, 238U, and the uranium mixture. Committed effective doses were re-assessed using BIOKMOD code by applying the new OIR excretion model and dose coefficients considering acute and/or chronic inhalation intake scenarios. A reduction by roughly a factor of four was obtained compared with former doses based on ICRP Publications 60/78/119, which is an important effect to take into consideration.

Extrapolation of Monte Carlo–Derived Noble Gas Air-Submersion Effective Dose Rate Coefficients for Selected Occupational Settings to Other Geometries and Radionuclides

External exposures to airborne radioactivity at nuclear power sites are typically based on semi-infinite clouds with uniform concentrations and corresponding effective dose rate coefficients (EDRCs), along with the simplifying assumption that the radioactive clouds extend to infinity around the receptor of interest. The two regulatory models that are typically employed for finite-cloud adjustments to submersion doses, namely, Regulatory Guide 1.183 for hemispherical clouds and the International Commission on Radiological Protection (ICRP) document ICRP-30 for spherical clouds, were purposely oversimplified to facilitate their implementation. As a result, dose projections can be significantly underestimated under certain circumstances, particularly with radionuclides emitting low-energy photons and/or particles. In addition, these adjustments do not account for scatter radiation off surrounding walls, ceilings, and floors of typical occupational settings. In recognition of these limitations and for the mitigation thereof, Veinot et al. published an article [Rad. and Environ. Biophy., Vol. 56, p. 453 (2017)] that provided monoenergetic photon, electron, and positron EDRCs based on elaborate Monte Carlo computations for submersion in three typical occupational settings at nuclear facilities (namely, an office, a laboratory, and a warehouse). Included in the article were also EDRCs for exposure to 45 noble gases airborne within the said occupational settings. However, extrapolation of the results to other settings was limited due to geometry variations. Even so, as described in the present paper, the Veinot et al. article provided the basis for the definition of a straightforward model for extrapolation of the Monte Carlo–derived EDRCs to other occupational settings and radionuclides. The objective of the EDRC extrapolation model in this paper is to provide a simple but comprehensive approach to licensing-basis submersion dose calculations consistent with the recommendations in ICRP Publication 103. The model is reasonably accurate and applicable to submersion volumes that are smaller than an office and larger than a typical warehouse, as would be needed for control room habitability evaluations. It is emphasized that the model is only suitable for the external dose computation of airborne noble gases and particulates; dose contributions via the inhalation pathway and from direct shine from contaminated surfaces need to be evaluated separately. The EDRC extrapolation model was programmed into an Excel workbook (OccuSetEDRCs-R0.xlsx) that is available to interested parties; see Supplementary Data section for further details. The model can handle all 1252 radionuclides in ICRP-107 and submersion volumes ranging between about 40 and 4000 m3, with an upper estimated error of about 10% on average for large submersion volumes.

Calculations of backpatch factors and conversion coefficients for the personal dose equivalent for X-ray beams used in radiodiagnostic

The purpose of this study is to use the Monte Carlo N-Particle radiation transport code (MCNP) to calculate the conversion coefficients of air kerma for personal dose equivalent (Hp [Sv]) and Back Scattering Factors (BSF) for X-ray beams used in diagnostic radiography at potentials of 40, 50, 60, 70, 80, 90, 100, 120, and 150 kVp. Additionally, effective dose is determined using male and female anthropomorphic voxel-based simulators (REX and REGINA) as recommended by the International Commission on Radiological Protection (ICRP) Publication 110, along with tissue weight factors from ICRP Publication 103 to simulate whole-body irradiation. The results indicate that the relationship between operational metrics such as Hp and risk limitation metrics like effective dose varies from 0.14 to 0.71 for Radiation Qualities in Radiation (RQR) beams from the X-ray source assembly and up to 0.64 for Radiation Qualities based on a phantom with an added aluminium filter (RQA). When comparing the obtained values for air kerma conversion coefficients for personal dose equivalent with the value adopted by the National Commission for Nuclear Energy (CNEN) of 1.14 Sv/Gy for individual dose determination, it is observed that at low energies, the values are overestimated by up to 44%, while at high energies, the values are underestimated by up to 67%.

National-level Dental Radiological Dose Management

There is growing public concern about patients' radiation exposure from recent medical imaging using ion-izing radiation. While medical radiation exposure has inevitable aspects, efforts to optimize doses in line with the principles of justification and optimization are necessary to maximize diagnostic value with the minimumradiation dose. Diagnostic reference levels have been proposed as a means to achieve optimization, in accordance with ICRP (International Commission on Radiological Protection) publications 60 and 73, as well as EC (European Commission) Directive 97/43/Euratom. According to Requirement 34 of the 'InternationalBasic Safety Standards' published by the IAEA in 2014, governments are recommended to establish national diagnostic reference levels. According to ICRP Publication 135, issued in 2017, national diagnostic reference levels should be revised regularly, preferably every 3-5 years, and more frequently in the case of technologicaladvancements, such as in the case of CBCT. In South Korea, diagnostic reference levels for various imagingmodalities have been established since 2007. In the field of dentistry, diagnostic reference levels for intraoral and panoramic radiography were established in 2009, and in 2018, diagnostic reference levels for dental CBCT were introduced for the first time. The purpose of this report is to discuss the current national-level patient radia-tion dose management in South Korea and explore its applications

Mathematical solutions in internal dose assessment: A comparison of Python-based differential equation solvers in biokinetic modeling

In biokinetic modeling systems employed for radiation protection, biological retention and excretion have been modeled as a series of discretized compartments representing the organs and tissues of the human body. Fractional retention and excretion in these organ and tissue systems have been mathematically governed by a series of coupled first-order ordinary differential equations (ODEs). The coupled ODE systems comprising the biokinetic models are usually stiff due to the severe difference between rapid and slow transfers between compartments. In this study, the capabilities of solving a complex coupled system of ODEs for biokinetic modeling were evaluated by comparing different Python programming language solvers and solving methods with the motivation of establishing a framework that enables multi-level analysis. The stability of the solvers was analyzed to select the best performers for solving the biokinetic problems. A Python-based linear algebraic method was also explored to examine how the numerical methods deviated from an analytical or semi-analytical method. Results demonstrated that customized implicit methods resulted in an enhanced stable solution for the inhaled 60Co (Type M) and 131I (Type F) exposure scenarios for the inhalation pathway of the International Commission on Radiological Protection (ICRP) Publication 130 Human Respiratory Tract Model (HRTM). The customized implementation of the Python-based implicit solvers resulted in approximately consistent solutions with the Python-based matrix exponential method (expm). The differences generally observed between the implicit solvers and expm are attributable to numerical precision and the order of numerical approximation of the numerical solvers. This study provides the first analysis of a list of Python ODE solvers and methods by comparing their usage for solving biokinetic models using the ICRP Publication 130 HRTM and provides a framework for the selection of the most appropriate ODE solvers and methods in Python language to implement for modeling the distribution of internal radioactivity.

Development of Respiratory Tract Organs for ICRP Pediatric Mesh-type Reference Computational Phantoms.

As part of the activities of the International Commission on Radiological Protection (ICRP) Task Group 103, the present study developed a new set of respiratory tract organs consisting of the extrathoracic, bronchial, bronchiolar, and alveolar-interstitial regions for newborn, 1-, 5-, 10-, and 15-y-old males and females for use in pediatric mesh-type reference computational phantoms. The developed respiratory tract organs, while preserving the original topologies of those of the pediatric voxel-type reference computational phantoms of ICRP Publication 143, have improved anatomy and detailed structure and also include μm-thick target and source regions prescribed in ICRP Publication 66. The dosimetric impact of the developed respiratory tract organs was investigated by calculating the specific absorbed fraction for internal electron exposures, which were then compared with the ICRP Task Group 96 values. The results showed that except for the alveolar-interstitial region as a source region, the pediatric mesh phantoms showed larger specific absorbed fractions than the Task Group 96 values. The maximum difference was a factor of ~3.5 for the extrathoracic-2 basal cell and surface as target and source regions, respectively. These results reflect the differences in the target masses and geometry caused by the anatomical enhancement of the pediatric mesh phantoms. For the alveolar-interstitial region as a source region, the pediatric mesh phantoms showed larger values for low energy ranges and lower values with increasing energies, owing to the differences in the size and shape of the alveolar-interstitial region.

Current Status of TRIPOLI-4® Monte Carlo Radiation Transport Code on Adult and Pediatric Computational Phantoms for Radiation Dosimetry Study

TRIPOLI-4® is a general-purpose Monte Carlo radiation transport code developed by the Service d’Études des Réacteurs et de Mathématiques Appliquées at CEA-Saclay. It uses continuous-energy nuclear data to simulate neutron, photon, electron, and positron transport in fields like radiation shielding, reactor physics, and nuclear criticality safety. To study radiation protection dosimetry in human tissues and organs, male and female adult computational phantoms from the Medical Internal Radiation Dose–Oak Ridge National Laboratory phantoms family and the International Commission on Radiological Protection (ICRP) publication 110 were recently modeled and calculated using the geometry options of TRIPOLI-4. To easily use the ICRP 110 voxel-based phantoms in different exposure scenarios, a newly developed phantom option is available in TRIPOLI-4 and its display tool T4G. This new phantom option is helpful for modeling one or more phantoms and for improving calculation performance in real irradiation environments. The 2020 published pediatric computational reference phantoms are accessible from ICRP publication 143. Male and female pediatric phantoms are also verified with the new T4G tool and TRIPOLI-4 code. This paper reports on recent works using TRIPOLI-4 on adult and pediatric computational phantoms. The modeling methods of stylized and voxel-based phantoms, the graphic displays of modeled phantoms with T4G, and the verification procedures for single-phantom and two-phantom application cases are presented. Validation for external and internal dosimetry calculations has been performed. Calculation results on organ dose S values for nuclear medicine applications are presented for single female and single male voxel phantoms using 131I and 177Lu radiation sources. Effective dose calculations for two-phantom cases using 99mTc and 18F sources are compared with traditional H*(10) calculations from nuclear medicine patient to patient caregiver.

MIRD Pamphlet No. 28, Part 2: Comparative Evaluation of MIRDcalc Dosimetry Software Across a Compendium of Diagnostic Radiopharmaceuticals.

Radiopharmaceutical dosimetry is usually estimated via organ-level MIRD schema-style formalisms, which form the computational basis for commonly used clinical and research dosimetry software. Recently, MIRDcalc internal dosimetry software was developed to provide a freely available organ-level dosimetry solution that incorporates up-to-date models of human anatomy, addresses uncertainty in radiopharmaceutical biokinetics and patient organ masses, and offers a 1-screen user interface as well as quality assurance tools. The present work describes the validation of MIRDcalc and, secondarily, provides a compendium of radiopharmaceutical dose coefficients obtained with MIRDcalc. Biokinetic data for about 70 currently and historically used radiopharmaceuticals were obtained from the International Commission on Radiological Protection (ICRP) publication 128 radiopharmaceutical data compendium. Absorbed dose and effective dose coefficients were derived from the biokinetic datasets using MIRDcalc, IDAC-Dose, and OLINDA software. The dose coefficients obtained with MIRDcalc were systematically compared against the other software-derived dose coefficients and those originally presented in ICRP publication 128. Dose coefficients computed with MIRDcalc and IDAC-Dose showed excellent overall agreement. The dose coefficients derived from other software and the dose coefficients promulgated in ICRP publication 128 both were in reasonable agreement with the dose coefficients computed with MIRDcalc. Future work should expand the scope of the validation to include personalized dosimetry calculations.

New skeletal dose coefficients of the ICRP-110 reference phantoms for idealized external fields to photons and neutrons using dose response functions (DRFs)

The International Commission on Radiological Protection (ICRP) Publication 116 was released to provide a comprehensive dataset of the dose coefficients (DCs) for external exposures produced with the adult reference voxel phantoms of ICRP Publication 110. Although an advanced skeletal dosimetry method for photons and neutrons using fluence-to-dose response functions (DRFs) was introduced in ICRP Publication 116, the ICRP-116 skeletal DCs were calculated by using the simple method conventionally used (i.e., doses to red bone marrow and endosteum approximated by doses to spongiosa and/or medullary cavities). In the present study, the photon and neutron DRFs were used to produce skeletal DCs of the ICRP-110 reference phantoms, which were then compared with the ICRP-116 DCs. For photons, there were significant differences by up to ∼2.8 times especially at energies <0.3 MeV. For neutrons, the differences were generally small over the entire energy region (mostly <20%). The general impact of the DRF-based skeletal DCs on the effective dose calculations was negligibly small, supporting the validity of the ICRP-116 effective DCs despite their skeletal DCs derived from the simple method. Meanwhile, we believe that the DRF-based skeletal DCs could be beneficial in better estimates of skeletal doses of individuals for risk assessments.

S-values estimation of positron-emitting radionuclides in the ICRP voxel-based adult male organs using a new Geant4-based code DoseCalcs: validation study.

Identifying the organs and tissues at risk from internal radiation exposure caused by radiopharmaceuticals requires determining the absorbed dose. The absorbed dose for radiopharmaceuticals is calculated by multiplying cumulated activity in source organs by the S-value, a crucial quantity that connects the energy deposited in the target organ and the emitting source one. It is defined as the ratio of absorbed energy in the target organ per unit of mass and unit of nuclear transition in the source organ. In this study, we used a new Geant4-based code called DoseCalcs to estimate the S-values for four positron-emitting radionuclides ([Formula: see text]C, [Formula: see text]N, [Formula: see text]O, and [Formula: see text]F) using decay and energy data from International Commission on Radiological Protection (ICRP) Publication 107. Twenty-three regions were simulated as radiation sources in the ICRP voxelized adult model developed in ICRP Publication 110. The Livermore physics packages were tailored to radionuclide photon mono-energy and [Formula: see text]-mean energy. The estimated S-values based on [Formula: see text]-mean energy show good agreement with those in the OpenDose data whose values were calculated using the full [Formula: see text] spectrum. The results provide new S-values data for selected source regions; hence, they could be used for comparison and adult-patient dose estimation.

Incorporation of micro-CT-based detailed bone models into ICRP adult mesh-type reference computational phantoms

Objective. The red bone marrow (RBM) and bone endosteum (BE), which are required for effective dose calculation, are macroscopically modeled in the reference phantoms of the international commission on radiological protection (ICRP) due to their microscopic and complex histology. In the present study, the detailed bone models were developed to simplify the dose calculation process for skeletal dosimetry. Approach. The detailed bone models were developed based on the bone models developed at the University of Florida. A new method was used to update the definition of BE region by storing the BE location indices using virtual sub-voxels. The detailed bone models were then installed in the spongiosa regions of the ICRP mesh-type reference computational phantoms (MRCPs) via the parallel geometry feature of the Geant4 code. Main results. Comparing the results between the detailed-bone-installed MRCPs and the original MRCPs with the absorbed dose to spongiosa and fluence-to-dose response function (DRF)-based methods, the DRF-based method showed much smaller but still significant differences. Compared with the values given in ICRP Publications 116 and 133, the differences were very large (i.e. several orders of magnitudes), due mainly to the anatomical improvement of the skeletal system in the MRCPs; that is, spongiosa and medullary cavity are fully enclosed by cortical bone in the MRCPs but not in the ICRP-110 phantoms. Significance. The detailed bone models enable the direct calculation of the absorbed doses to the RBM and BE, simplifying the dose calculation process and potentially improving the consistency and accuracy of skeletal dosimetry.

On the operational quantity for eye lens neutron dosimetry considering ICRU 95 report

For planned occupational exposure situations, the International Commission on Radiological Protection (ICRP) publication 118 recommends an equivalent dose limit for the lens of the eye of 20 mSv yr−1 averaged over 5 yr with no single year exceeding 50 mSv. 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. As compliance with the eye-lens dose limit will be based on the summation of doses received from all types of radiation, applicable to a variety of workplaces, the contribution of neutrons to eye lens dose will be important where it contributes a significant fraction of the total dose to the eye lens. This work presents and discusses computed personal absorbed dose (D lens/Φ), and personal dose equivalent (H p(3)/Φ) as well as a newly proposed relative biological effectiveness (RBE)-weighted absorbed dose (RBE × D lens/Φ) conversion coefficients for the lens of the eye for neutron exposure at incident energies from thermal to ∼20 MeV. The D lens/Φ coefficients were obtained from a simulation model developed for this study that contains the stylised eye model embedded in the adult UF-ORNL mathematical phantom. The modelling techniques used in these simulations were also used to calculate H p(3)/Φ for the International Commission on Radiation Units and Measurements (ICRU) slab and cylinder phantoms. All simulations carried out for this study utilised the Monte Carlo N-Particle (MCNP) series of codes. The results are compared with the related published data. The issue of compliance with the current equivalent dose limit for the lens of the eye is addressed from a neutron perspective considering the recent proposed redefinition of the operational quantities for external radiation exposure in ICRU report 95. The use of a radiation weighted absorbed dose (RBE × D lens, in Gy) is proposed for the tissue reactions in the eye-lens for neutron radiation as per the National Council on Radiation Protection and Measurements report 180, and in line with the recent review and revision of the System of Radiological Protection To Keeping the ICRP Recommendations Fit for Purpose, which states that RBE weighted dose should be used for high-Linear energy transfer (LET) radiations such as neutrons. This confirms the earlier statement in ICRP publication 92, paragraph 297 and reiterated in the Executive summary, paragraph (q) of ICRP publication 118. The proposed approach would provide an operational quantity consistent with the units of the new eye-lens dose limits without being overly conservative.

Development of alimentary tract organs for ICRP pediatric mesh-type reference computational phantoms

In line with the activities of Task Group 103 under the International Commission on Radiological Protection (ICRP), the present study was conducted to develop a new set of alimentary tract organs consisting of the oral cavity, oesophagus, stomach, small intestine, and colon for the newborn, 1 year-old, 5 year-old, 10 year-old, and 15 year-old males and females for use in the pediatric mesh-type reference computational phantoms (MRCPs). The developed alimentary tract organs of the pediatric MRCPs, while nearly preserving the original topology and shape of those of the pediatric voxel-type reference computational phantoms (VRCPs) of ICRP Publication 143, present considerable anatomical improvement and include all micrometre-scale target and source regions as prescribed in ICRP Publication 100. To investigate the dosimetric impact of the developed alimentary tract organs, organ doses and specific absorbed fractions were computed for certain external exposures to photons and electrons and internal exposures to electrons, respectively, which were then compared with the values computed using the current ICRP models (i.e. pediatric VRCPs and ICRP-100 stylised models). The results showed that for external exposures to penetrating radiations (i.e. photons >0.04 MeV), there was generally good agreement between the compared values, within a 10% difference, except for the oral mucosa. For external exposures to weakly penetrating radiations (i.e. low-energy photons and electrons), there were significant differences, up to a factor of ∼8300, owing to the geometric difference caused by the anatomical enhancement in the MRCPs. For internal exposures of electrons, there were significant differences, the maximum of which reached a factor of ∼73 000. This was attributed not only to the geometric difference but also to the target mass difference caused by the different luminal content mass and organ shape.

IDAC-Bio, A Software for Internal Dosimetry Based on the New ICRP Biokinetic Models and Specific Absorbed Fractions.

Radiation dosimetry is central to virtually all radiation safety applications, optimization, and research. It relates to various individuals and population groups and to miscellaneous exposure situations—including planned, existing, and emergency situations. The International Commission on Radiological Protection (ICRP) has developed a new computational framework for internal dose estimations. Important components are more detailed and improved anatomical models and more realistic biokinetic models than before. The ICRP is currently producing new organ dose and effective dose coefficients for occupational intakes of radionuclides (OIR) and environmental intakes of radionuclides (EIR), which supersede the earlier dose coefficients in Publication 68 and the Publication 72 series, respectively. However, the ICRP only publishes dose coefficients for a single acute intake of a radionuclide and for an integration period of 50 years for intake by adults and to age 70 years for intakes by pre-adults. The new software, IDAC-Bio, performs committed absorbed dose and effective dose calculations for a selectable intake scenario, e.g., for a continuous intake or an intake during x hours per day and y days per week, and for any selected integration time. The software uses the primary data and models of the ICRP biokinetic models and numerically solves the biokinetic model and calculates the absorbed doses to organs and tissues in the ICRP reference human phantoms. The software calculates absorbed dose using the nuclear decay data in ICRP publication 107. IDAC-Bio is a further development and an important addition to the internal dosimetry program IDAC-Dose2.1. The results generated by the software were validated against published ICRP dose coefficients. The potential of the software is illustrated by dose calculations for a nuclear power plant worker who had been exposed to varying levels of 60Co and who had undergone repeated whole-body measurements, and for a hypothetical member of the public subject to future releases of 148Gd from neutron spallation in tungsten at the European Spallation Source.

Radiation detriment calculation methodology: summary of ICRP Publication 152

Radiation detriment is a concept to quantify the burden of stochastic effects from exposure of the human population to low-dose and/or low-dose-rate ionising radiation. As part of a thorough review of the system of radiological protection, the International Commission on Radiological Protection (ICRP) has compiled a report on radiation detriment calculation methodology as Publication 152. It provides a historical review of the detriment calculation with details of the procedure used in ICRP Publication 103. A selected sensitivity analysis was conducted to identify the parameters and calculation conditions that can be major sources of variation and uncertainty. It has demonstrated that sex, age at exposure, dose and dose-rate effectiveness factor, dose assumption in the lifetime risk calculation, and lethality fraction have a substantial impact on the calculated values of radiation detriment. Discussions are also made on the issues to be addressed and possible ways for improvement toward the revision of general recommendations. These include update of the reference population data and cancer severity parameters, revision of cancer risk models, and better handling of the variation with sex and age. Finally, emphasis is placed on transparency and traceability of the calculation, along with the need to improve the way of expressing and communicating the detriment.

A Modification to the Situation-based Scheme for Sorting Exposures Proposed in ICRP Publication 103.

The International Commission on Radiological Protection (ICRP) issued its current recommendations in 2007, introducing an evolution from a previously process-based protection approach (practices and interventions) to a situation-based approach (planned, existing, and emergency). In most countries, radiation protection standards are generally in line with the new recommendations by the ICRP. Unfortunately, in many cases, situations involving sources of natural or artificial origin may be in a state of contradiction. Logical incompleteness and unsoundness in the framework of exposure situations can be found in current ICRP recommendations. There are situations that do not readily fall into any category. To contribute to a more reasonable situation-based approach for the radiological protection community, we recommend here a possible framework of exposure situations, i.e., planned, unplanned, and naturally occurring, that forms one coherent system. We believe this system will provide increased clarity that will help resolve possible sorting problems.

Сравнительный анализ оценки радиационных рисков медицинского диагностического облучения пациентов на основе метода Международной комиссии по радиологической защите и методических рекомендаций Роспотребнадзора

This work is devoted to a comparative analysis of the assessment of radiation risks of medical diagnostic exposure of patients using a method based on the recommendations of the International Commission on Radiological Protection (ICRP) and the method presented in the methodological recommendations of Rospotrebnadzor (MR 2.6.1.0215-20). In the course of the study, the authors analyzed the results of calculations of the lifetime attributable risk of oncological mortality calculated by these methods. As an example, the procedure of a single passage of computed tomography of the chest organs without contrast was considered. In the first case, the risk value was calculated using the methodology of ICRP Publication 103 based on organ doses for the following ages of men and women: examination at 22, 27, 32, 37, 42, 47, 52, 57, 62, 67, 72, 77, 82 and at 87 years old. In the second case, the data given in the tables presented in the methodological recommendations were considered, but without considering hereditary effects. The analysis was carried out for two cases: when the difference between effective doses between computed tomography procedures did not exceed 30%, and when the difference was more than 30%. The authors of the work found that the difference in the risk assessment between the two methods exceeds 1.5 times for the age groups of 65 years and older, for men and women, if the effective doses are close. If the difference in dose between the two procedures was more than 30%, then a 1.5-fold difference in risks was observed in age groups 60 years and older. Risk assessment was performed with account of age and gender characteristics, as well as with the use of medical and demographic data of the Russian population.

Keeping the ICRP recommendations fit for purpose

The International Commission on Radiological Protection (ICRP) has embarked on a review and revision of the system of Radiological Protection that will update the 2007 general recommendations in ICRP Publication 103. This is the beginning of a process that will take several years, involving open and transparent engagement with organisations and individuals around the world. While the system is robust and has performed well, it must adapt to address changes in science and society to remain fit for purpose. The aim of this paper is to encourage discussions on which areas of the system might gain the greatest benefit from review, and to initiate collaborative efforts. Increased clarity and consistency are high priorities. The better the system is understood, the more effectively it can be applied, resulting in improved protection and increased harmonisation. Many areas are identified for potential review including: classification of effects, with particular focus on tissue reactions; reformulation of detriment, potentially including non-cancer diseases; re-evaluation of the relationship between detriment and effective dose, and the possibility of defining detriments for males and females of different ages; individual variation in the response to radiation exposure; heritable effects; and effects and risks in non-human biota and ecosystems. Some of the basic concepts are also being considered, including the framework for bringing together protection of people and the environment, incremental improvements to the fundamental principles of justification and optimisation, a broader approach to protection of individuals, and clarification of the exposure situations introduced in 2007. In addition, ICRP is considering identifying where explicit incorporation of the ethical basis of the system would be beneficial, how to better reflect the importance of communications and stakeholder involvement, and further advice on education and training. ICRP invites responses on these and other areas relating to the review of the System of Radiological Protection.