Internal Dose Assessment Research Articles

MIRD Pamphlet No. 30: MIRDfit-A Tool for Fitting of Biodistribution Time-Activity Data for Internal Dosimetry.

In nuclear medicine, estimating the number of radioactive decays that occur in a source organ per unit administered activity of a radiopharmaceutical (i.e., the time-integrated activity coefficient [TIAC]) is an essential task within the internal dosimetry workflow. TIAC estimation is commonly derived by least-squares fitting of various exponential models to organ time-activity data (radiopharmaceutical biodistribution). Rarely, however, are methods used to objectively determine the model that best characterizes the data. Additionally, the uncertainty associated with the resultant TIAC is generally not evaluated. As part of the MIRDsoft initiative, MIRDfit has been developed to offer a biodistribution fitting software solution that provides the following essential features and advantages for internal dose assessment: nuclear medicine-appropriate fit functions; objective metrics for guiding best-fit selection; TIAC uncertainty calculation; quality control and data archiving; integration with MIRDcalc software for dose calculation; and a user-friendly Excel-based interface. For demonstration and comparative validation of MIRDfit's performance, TIACs were derived from serial imaging studies involving 18F-FDG and 177Lu-DOTATATE using MIRDfit. These TIACs were then compared with TIAC estimates obtained using other software. In most cases, the TIACs agreed within approximately 10% between MIRDfit and the other software. MIRDfit has been endorsed by the MIRD Committee of the Society of Nuclear Medicine and Molecular Imaging and has been integrated into the MIRDsoft suite of free dosimetry software; it is available for download at no user cost (https://mirdsoft.org/).

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.

Determination of Activity Concentrations and Internal Dose Assessments in Urine from Workers in Iraqi Nuclear Research Center, Iraq

To detect tritium levels in the human body, tritium in the urine must be measured. The investigation of the level of tritium activity in the urine of workers of the Iraqi Nuclear Research Center (INRC) has been realized in order to assess vocational exposure levels. Three worker groups in INRCs were chosen, involved in the Central Laboratories Directorate (CLD), the Decommissioning Directorate (DD), and the Radioactive Waste Treatment and Radioactive Waste Store Directorate (WTWSD), among others in Al-Tuwiatha site, in addition to the control group, as the research targets. Tritium concentration was determined in urine samples from 52 workers (aged 32-61) and 13 control people (aged 30-58). 65 urine samples of 50 mL were collected and analyzed. The Eichrom's Tritium Column way was used to handle the samples of urine, and the distillate tritium activity concentration was then assessed using Liquid Scintillation Counter. The effective and annual effective doses were estimated for the internal doses. The accuracy of urine sample measurements had been tested. The highest level of tritium activity samples of urine from three groups of workers in INRCs is 65.401 Bq/L, and for control group is 39.300 Bq/L, activity of tritium concentrations estimated by using counting scintillation of liquid (LSC) were found to be lower than Minimum Detectable Activity (MDA) of 1.95 Bq/L. The effective and annual effective dose from tritium was also assessed depend on results measured previously and standard values adapted by the International Commission on Radiological Protection. The effective highest doses for workers groups and control group were 497.048 ×10-3 and 298.680×10-3 nSv respectively. Highest annual effective dose 5757.773×10-3 and 5167.042×10-3 nSv respectively. These results were lowest than the allowed effective dose for the workers.

Modeling Plutonium Decorporation in a Female Nuclear Worker Treated with Ca-DTPA after Inhalation Intake.

The present work models plutonium (Pu) biokinetics in a female former nuclear worker. Her bioassay measurements are available at the US Transuranium and Uranium Registries. The worker was internally exposed to a plutonium-americium mixture via acute inhalation at a nuclear weapons facility. She was medically treated with injections of 1 g Ca-DTPA on days 0, 5, and 14 after the intake. Between days 0 and 20, fecal and urine samples were collected and analyzed for 239Pu and 241Am. Subsequently, she was followed up for bioassay monitoring over 14 y, with additional post-treatment urine samples collected and analyzed for 239Pu. The uniqueness of this dataset is due to the availability of: (1) both early and long-term bioassay data from a female with plutonium intake; (2) data on chelation therapy for a female; and (3) fecal measurement results. Chelation therapy with Ca- and/or Zn-salts of DTPA is known to aid in reducing the internal radiation dose by enhancing the excretion of plutonium and americium from the body. Such enhancement affects plutonium biokinetics in the human body, posing a challenge to the internal dose assessment. The current radiation dose assessment practice is to exclude the data affected by Ca-DTPA from the analysis. The present analysis is the first to explicitly model the chelation-affected bioassay data in a female by using a newly developed chelation model. Thus, the bioassay data collected during and after the Ca-DTPA administrations were used for biokinetic modeling and dose assessment. The Markov Chain Monte Carlo method was used to investigate model parameter uncertainty, based on the bioassay data and assumed prior probability distributions. A χ2/nData (number of data points) ≈ 1 was observed in this study, which indicates self-consistency of the data with the model. Results of this study show that the worker's 239Pu intake was 12 Bq, with a committed effective dose to the whole-body of 1.2 mSv and a committed equivalent dose to the bone surfaces, liver, and lungs of 37.8, 9.1, and 0.8 mSv, respectively. This study also discusses the worker's dose reduction due to chelation treatment.

Improvement of internal exposure assessments of the inhalation of fuel-type hot particles during long-term outages

During outages at nuclear power plants, much more care for radiation workers against internal exposure should be ensured given that more hot particles exist relative to the amount during normal operation. If fuel-type hot particles (FTHP) are inhaled, they can cause more severe health risks compared to activation-type hot particles (ATHP), which contain 60Co, due to the alpha-emitting nuclides within FTHPs. The activities of difficult-to-measure nuclides within FTHPs inhaled by workers are inferred by the age-dating technique using a141Ce/144Ce ratio as measured by whole-body counters. However, this method may be limited to outages that last for only a few months due to the short half-life (32.5 days) of 141Ce. We studied the feasibility of utilizing 241Am, a nuclide with a long half-life of 432.6 years, as an alternative to 141Ce. Additionally, we improved the performance of a stand-type whole-body counter for low-energy gamma spectroscopy to meet the criterion (RMSE ≤0.25) specified in ANSI/HPS N13.30–2011 by employing an artificial neural network (ANN). This study can contribute to more rapid and accurate internal dose assessments for workers who have inhaled FTHPs during long-term outages at nuclear power plants.

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.

Tritium concentration in bottled drinking water and internal dose assessment

Tritium concentration in bottled drinking water and internal dose assessment

Validation of automated image co-registration integrated into in-house software for voxel-based internal dosimetry on single-photon emission computed tomography images.

To develop an automated co-registration system and test its performance, with and without a fiducial marker, on single-photon emission computed tomography (SPECT) images. Three SPECT/CT scans were acquired for each rotation of a Jaszczak phantom (to 0°, 5°, and 10° in relation to the bed axis), with and without a fiducial marker. Two rigid co-registration software packages-SPM12 and NMDose-coreg-were employed, and the percent root mean square error (%RMSE) was calculated in order to assess the quality of the co-registrations. Uniformity, contrast, and resolution were measured before and after co-registration. The NMDose-coreg software was employed to calculate the renal doses in 12 patients treated with 177Lu-DOTATATE, and we compared those with the values obtained with the Organ Level INternal Dose Assessment for EXponential Modeling (OLINDA/EXM) software. The use of a fiducial marker had no significant effect on the quality of co-registration on SPECT images, as measured by %RMSE (p = 0.40). After co-registration, uniformity, contrast, and resolution did not differ between the images acquired with fiducial markers and those acquired without. Preliminary clinical application showed mean total processing times of 9 ± 3 min/patient for NMDose-coreg and 64 ± 10 min/patient for OLINDA/EXM, with a strong correlation between the two, despite the lower renal doses obtained with NMDose-coreg. The use of NMDose-coreg allows fast co-registration of SPECT images, with no loss of uniformity, contrast, or resolution. The use of a fiducial marker does not appear to increase the accuracy of co-registration on phantoms.

CADORmed: a tool for internal dose assessment.

CADORmed is a free bespoke Excel® tool for committed effective dose assessment using latest dose coefficients from ICRP OIR publications. The field of application of CADORmed is special monitoring, and it is not available for the dose assessment of chronic exposure. Calculations are made according to EURADOS guidelines and principles (EURADOS report 2013-1). The Chi-squared test for the goodness of fit is made with a scattering factor for type A and type B errors according to the EURADOS report. The Intake is calculated with the maximum likelihood method. Measurements that are below the detection limit are incorporated by the use of an allocated value equal to one-half or one-quarter of the detection limit. The Identification of rogue data can easily be achieved. Advanced options may also be used: mixed ingestion and inhalation, mixture of default absorption types, correction for DTPA treatment, calculation with a new intake and adjustment when the date of intake are unknown. The validation of the tool has been included in the work plan of EURADOS WG 7. The validation plan has been defined and the validation tests completed. All changes are traced in a Quality Assurance document.

Determination of internal gamma dose assessment due to natural radioactivity for river sand using a gamma-ray spectrometer.

River sand is an environmental component that is always used as a building material in India and hence the present study, activity concentrations of 226Ra, 232Th and 40K have been measured in sand samples collected from Ponnai river, Tamil Nadu using a high-resolution gamma-ray spectrometer consisting of high-purity germanium detector. Calculated mean specific activity is 31, 84 and 416Bqkg-1 for 226Ra, 232Th and 40K, respectively. These results reveal that 226Ra was found to be less than the world average value of 33Bqkg-1 whereas 232Th and 40K were higher than the global mean value of 30 and 400Bqkg-1, respectively. In order to assess the internal dose to population a standard index radium equivalent activity (Raeq) is calculated for these samples. Based on the obtained results, it is seen that these sand samples do not pose significant health hazards to the inhabitants of the houses constructed.

In Vivo evaluation of newly synthesized 213Bi-conjugated alpha-melanocyte stimulating hormone (α-MSH) peptide analogues in melanocortin-1 receptor (MC1-R) positive experimental melanoma model

Given the rising pervasiveness of melanocortin-1 receptor (MC1-R) positive melanoma malignum (MM) and pertinent metastases, radiolabelled receptor-affine alpha-melanocyte stimulating hormone-analogue (α-MSH analogue) imaging probes would be of crucial importance in timely tumor diagnostic assessment. Herein we aimed at investigating the biodistribution and the MM targeting potential of newly synthesized 213Bi-conjugated MC1-R specific peptide-based radioligands with the establishment of MC1-R overexpressing MM preclinical model. DOTA-conjugated NAP, -HOLD, -FOLD, -and MARSamide were labelled with 213Bi. Ex vivo biodistribution studies were conducted post-administration of 3.81 ± 0.32 MBq [213Bi]Bi-DOTA conjugated deriva-tives into twenty B16-F10 tumor-bearing C57BL/6 J and healthy mice. Organ Level Internal Dose Assessment (OLINDA) and IDAC-Dose were used to calculate translational data-based absorbed radiation dose in human organs. Moderate or low %ID/g uptake of [213Bi]Bi-DOTA conjugated NAP, -HOLD, -and MARSamide and significantly increased [213Bi]Bi-DOTA-FOLDamide accumulation was observed in the thoracic and abdominal organs (p ≤ 0.01). High [213Bi]Bi-DOTA-NAP (%ID/g:3.76 ± 0.96), -and FOLDamide (%ID/g:3.28 ± 0.95) tumor tracer activity confirmed their MC1-R-affinity. The bladder wall received the highest radiation absorbed dose followed by the kidneys (bladder wall: 1.95·10-2 and 8.97·10-2 mSv/MBq; kidneys: 7.47·10-3 vs. 5.88·10-2 mSv/MBq measured by IDAC and OLINDA; respectively) indicating the suitability of the NAPamide derivative for clinical use. These novel [213Bi]Bi-DOTA-linked peptide probes displaying meaningful MC1-R affinity could be promising molecular probes in MM imaging.

Reproductive and developmental toxicity following exposure to organophosphate ester flame retardants and plasticizers, triphenyl phosphate and isopropylated phenyl phosphate, in Sprague Dawley rats.

Two organophosphate esters used as flame retardants and plasticizers, triphenyl phosphate (TPHP) and isopropylated phenyl phosphate (IPP), have been detected in environmental samples around the world. Human exposure primarily occurs via oral ingestion with reported higher concentrations in children. Currently, there are no data to evaluate potential risk from exposure to either TPHP or IPP during fetal development. These short-term perinatal studies in rats provide preliminary toxicity data for TPHP and IPP, including information on transfer to fetus/offspring and across the pup blood-brain barrier. In separate experiments, TPHP or IPP were administered via dosed feed at concentrations 0, 1000, 3000, 10 000, 15 000, or 30 000 ppm to time-mated Hsd:Sprague Dawley SD rats from gestation day (GD) 6 through postnatal day (PND) 28; offspring were provided dosed feed at the same concentration as their dam (PND 28-PND 56). TPHP- and IPP-related toxicity resulted in removal of both 30 000 ppm groups on GD 12 and 15 000 ppm IPP group after parturition. Body weight and organ weights were impacted with exposure in remaining dams. Reproductive performance was perturbed at ≥10 000 ppm TPHP and all IPP exposure groups. In offspring, both TPHP- and IPP-related toxicity was noted in pups at ≥10 000 ppm as well as reduction in bodyweights, delays in pubertal endpoints, and/or reduced cholinesterase enzyme activity starting at 1000 ppm TPHP or IPP. Preliminary internal dose assessment indicated gestational and lactational transfer following exposure to TPHP or IPP. These findings demonstrate that offspring development is sensitive to 1000 ppm TPHP or IPP exposure.

Development of Self-Questionnaire for Internal Dose Assessment by Food Ingestion

Background: The accident at the Fukushima Daiichi nuclear power plant increased the level of anxiety related to the radioactive contamination of various foods sourced in Japan. Particularly, after the accident, the detection of artificial radionuclides in locally produced foods raised food safety concerns. In this study, the radioactivity concentrations and annual ingestions of <sup>40</sup>K and <sup>137</sup>Cs in food products commonly and frequently consumed by the general public were investigated, and the annual effective dose of each was evaluated.Materials and Methods: The 2016–2018 data from the Radiation Safety Management Report released by the Korea Nuclear Safety Technology Center was referenced for the evaluation of the amounts of <sup>40</sup>K and <sup>137</sup>Cs contained in food. Using the food-ingestion survey mentioned above as a reference, we selected 62 foods to include in our radioactivity concentration and dose assessment. We also developed a questionnaire and evaluated the responses from the subjects who answered the questionnaire.Results and Discussion: The radioactivity concentration of <sup>137</sup>Cs was found to be close to or below the level of minimum detectable activity. Additionally, the annual ingestion of 62 foods was 294.77 kg/yr, the effective doses from <sup>40</sup>K and <sup>137</sup>Cs were 136.4 and 0.163 μSv/yr, respectively.Conclusion: Thus, the findings confirmed that the effective dose from <sup>40</sup>K and <sup>137</sup>Cs in food tends to be lower than the effective dose limit of 1 mSv/yr suggested by the International Commission on Radiological Protection (ICRP) Publication 60. The questionnaire developed in this study is expected to be useful for estimating the annual effective dose status of Korean adults who consume foods containing <sup>40</sup>K and <sup>137</sup>Cs.

Determination of counting efficiency considering the biodistribution of 131I activity in the whole-body counting measurement

Whole-body counters are widely used to assess internal contamination after a nuclear accident. However, it is difficult to determine radioiodine activity due to limitations in conventional calibration phantoms. Inhaled or ingested radioiodine is heterogeneously distributed in the human body, necessitating time-dependent biodistribution for the assessment of the internal contamination caused by the radioiodine intake. This study aims at calculating counting efficiencies considering the biodistribution of 131I in whole-body counting measurement. Monte Carlo simulations with computational human phantoms were performed to calculate the whole-body counting efficiency for a realistic radioiodine distribution after its intake. The biodistributions of 131I for different age groups were computed based on biokinetic models and applied to age- and gender-specific computational phantoms to estimate counting efficiency. After calculating the whole-body counting efficiencies, the efficiency correction factors were derived as the ratio of the counting efficiencies obtained by considering a heterogeneous biodistribution of 131I over time to those obtained using the BOMAB phantom assuming a homogeneous distribution. Based on the correction factors, the internal contamination caused by 131I can be assessed using whole-body counters. These correction factors can minimize the influence of the biodistribution of 131I in whole-body counting measurement and improve the accuracy of internal dose assessment.

A method for the calculation of annual limits on intake

Abstract The German Radiation Protection Ordinance issued in 1989 – as well as ICRP Publication 30 – included tables of annual limits on intake (ALIs). These are secondary limits derived to indicate those intake values which are not allowed to be exceeded within a year to meet the primary annual dose limits. The latest ICRP Publications dealing with doses for workers give no ALIs any more and also the IAEA Basic Safety Standards and the Council Directive of the European Commission do not list such values. And also the new Radiation Protection Ordinance as well as the federal gazette with dose coefficients do not include such values any more. However, it may be that they will be included as guiding levels in the calculation principles for the assessment of internal doses. This paper will give reasons why there are no ALIs given any more, it will show how ALIs could be calculated, it will show the influence of additional dose restrictions for female workers of child-bearing age, and it will discuss if radiation protection for the unborn child would be fulfilled if the dose limits of the radiation protection ordinance are met.

APPLICABILITY OF A HANDHELD LaBr3(Ce) SPECTROMETER FOR INTERNAL RADIOACTIVE CONTAMINATION EXAMINATION IN RADIOLOGICAL EMERGENCY

In case of a nuclear or radiological emergency, there may be a very large population of individuals being affected by radiation exposure. Rapid and on-site examinations of possible internal radioactive contaminations are required for early dose assessment and large-scale screening. With the appropriate methodology, early dose information of internal exposure can be instantly obtained by a handheld spectrometer only. In this study, we extended the use of a handheld LaBr3 spectrometer to rough internal dose assessment. A family of real source BOMAB phantoms was applied for efficiency calibration of different detecting geometries. Detecting limits of several nuclides of major concern was also investigated. The result of this study can be used for initial dose assessment and medical triage during the first response of nuclear and radiological emergencies.

IDACstar2.0: A MCNP application to perform realistic dose estimations from internal or external contamination with radiopharmaceuticals

The IDACstar (Internal Dose Assessment by Computer) software is a Monte Carlo based standalone MCNP input generator software developed to perform complex radiation dose calculations using both the paediatric and adult ICRP reference phantoms. An extension of the IDACstar software was applied to two different exposure scenarios. One of the scenarios was a hypothetical case of a contaminated diaper from urine of a 1-year old patient undergoing a diagnostic kidney examination with 99mTc-labelled mercaptoacetyl triglycerine (MAG3), which resulted in an effective dose estimate of 0.088 mSv to the child. The second case was a clinical case of 18F-FDG extravasation in an adult patient and applied to all 10 age-specific phantoms with weight-modified administrations of the radiopharmaceutical. The effective dose was estimated for the 15-, 10-, 5- and 1- year old as 5.1 mSv, 4.9 mSv, 4.7 mSv, and 4.7 mSv, respectively. The updated version of the IDACstar software provides the possibility of performing realistic dose estimations to paediatric patients from internal or external contamination with radiopharmaceuticals.

Dose Coefficients for Internal Dose Assessments for Exposure to Radioactive Fallout.

This paper presents values as well as the bases for calculating internal dose coefficients suitable for estimating organ doses from the exposure to radioactive fallout that could result from the detonation of a nuclear fission device. The 34 radionuclides discussed are the same as those given in a priority list of radionuclides for fallout dose assessments presented in a companion overview paper. The radionuclides discussed are those that are believed to account for a preponderance of the organ doses that might be received by intake by persons of all ages (including in utero and via breast feeding for infants) following exposure to radioactive fallout. The presented dose coefficients for ingestion account for age and include modifications for variations in solubility with distance as discussed previously in the literature, and those for inhalation similarly account for age, solubility, and particle sizes that would be relevant at various distances of exposure as discussed in a companion paper on ingestion dose methods. The proposed modifications peculiar to radioactive fallout account for systematic changes in solubility and particle sizes with distance from the site of detonation, termed here as the region of "local fallout" and the region "beyond local fallout." Brief definitions of these regions are provided here with more detailed discussion in a companion paper on estimating deposition of fallout radionuclides. This paper provides the dose coefficients for ingestion and inhalation (for particle sizes of 1 μm, 5 μm, 10 μm, and 20 μm) for the region "local fallout." These dose coefficients for "local fallout" are specific for particles formed in a nuclear explosion that can be large and have radionuclides, particularly the more refractory ones, distributed throughout the volume where the radionuclide has reduced solubility. The dose coefficients for the region "beyond local fallout" are assumed to be the ones published by the International Commission on Radiological Protection (ICRP) in 1995. Comparisons of the presented dose coefficients are made with values published by the ICRP.

Age-Specific Thyroid Internal Dose Estimation for Koreans

Background: The International Commission on Radiological Protection is preparing to provide reference dose coefficients for environmental radioiodine intake based on newly developed age-specific biokinetic models. However, the biokinetics of iodine has been reported to be strongly dependent on the dietary intake of stable iodine; for example, the thyroidal uptake of iodine may be substantially lower in iodine-rich regions than in iodine-deficient regions. Therefore, this study attempted to establish a system of age-specific thyroid dose estimation for South Koreans, whose daily iodine intakes are significantly higher than that of the world population.Materials and Methods: Korean age-specific biokinetic parameters and thyroid masses were derived based on the previously developed Korean adult model and the Korean anatomical reference data for adults, respectively. This study complied with the principles used in the development of age-specific biokinetic models for world population and used the ratios of baseline values for each age group relative to the value for adults to derive age-specific values.Results and Discussion: Biokinetic model predictions based on the Korean age-specific parameters showed significant differences in iodine behaviors in the body compared to those predicted using the model for the world population. In particular, the Korean age-specific thyroid dose coefficients for 129I and 131I were considerably lower than those calculated for the world population (25%–76% of the values for the world population).Conclusion: These differences stress the need for Korean-specific internal dose assessments for infants and children, which can be achieved by using the data calculated in this study.

Assessment of absorbed dose for Zr-89, Sm-153 and Lu-177 medical radioisotopes: IDAC-Dose2.1 and OLINDA experience

ObjectiveIn this article, IDAC-Dose2.1 and OLINDA computer codes are compared as they are the most widely used software tools for internal dosimetry assessment at the present time. OLINDA/EXM personal computer code was created as a replacement for the widely used MIRDOSE3.1 code. IDAC-Dose2.1 was developed based on the ICRP specific absorbed fractions and computational framework of internal dose assessment given for reference adults in ICRP Publication 133. IDAC uses cumulated activities per administered activity in hours and calculates the absorbed dose and the effective dose. The program calculates the dose in the Eckerman stylized family phantoms. It is useful in standardizing and automating internal dose calculations, assessing doses in clinical trials with radiopharmaceuticals, making theoretic calculations for existing pharmaceuticals, teaching, and other purposes. MethodsTo produce such a comparison, the results of this work were compared with available published data in the literature on radiopharmaceuticals. Radiopharmaceuticals with 89Zr, 153Sm, 177Lu radionuclides are used as the basis for the comparison. 89Zr, 153Sm, 177Lu radionuclides are regarded as the future of radiopharmaceutical treatment. For 89Zr, two different labelled carriers, Zr-89_cMAb U36 and Zr-89 Panitumumab, were used on patients. ResultsThe results show a clear difference in terms of absorbed dose of the Zr-89 radiopharmaceuticals for red bone marrow when calculated by IDAC-Dose2.1 (0.76 mGy/MBq), while the estimated absorbed dose in literature results is 0.07 mGy/MBq and 0.14 mGy/MBq when the calculation is done by OLINDA program. In the case of 177Lu-EDTMP, the absorbed dose in red bone marrow is in reasonable agreement (0.63 mGy/MBq and 0.8 mGy/MBq for IDAC-Dose2.1 and OLINDA, respectively). A significant difference was found for the absorbed dose in the bone surface, which was almost twice as high for OLINDA (2.1 mGy/MBq for IDAC-Dose2.1 and 5.4 mGy/MBq for OLINDA). In some direct cases, the calculated absorbed dose in the urinary bladder wall with OLINDA is ten times higher compared to WinAct (which was utilized to calculate the total activity in the organs and tissues) and IDAC 2.1. These results are considered key to greater accuracy in internal dose calculation.