Medical Internal Radiation Dose Research Articles

Evaluation of 68Ga-PSMA-11 PET/CT: a Phase 1 clinical study in Japanese patients with primary, recurrent, or suspected recurrent prostate cancer.

Prostate-specific membrane antigen (PSMA)-targeted radiopharmaceuticals allow whole-body imaging to detect prostate cancer (PC). Positron emission tomography imaging using gallium-68 (68Ga)-PSMA-11 has been shown to have a favorable safety and tolerability profile and high diagnostic performance. The study evaluates the safety and pharmacokinetics of 68Ga-PSMA-11 in Japanese patients with primary, recurrent, or suspected recurrent prostate cancer. This single arm study enrolled Japanese patients with primary PC (n = 3), suspected recurrent PC following radical prostatectomy (n = 4), or suspected recurrent PC following radical radiotherapy (n = 3). All patients received a single intravenous dose of 68Ga-PSMA-11 2.0MBq/kg (±10%) followed by PSMA PET imaging and safety and pharmacokinetic evaluations. Based on the blood concentrations of 68Ga-PSMA-11 and the radioactivity distribution rate in each organ/tissue, the absorbed doses in major organs/tissues and the whole-body effective dose were calculated by the Medical Internal Radiation Dose method. Ten patients were enrolled. Mean age was 73.3 ± 4.8years, and median prostate-specific antigen was 8.250 ng/mL. Five patients (50%) experienced a total of 6 adverse events, and no grade ≥ 2 adverse events or serious adverse events were reported. No clinically significant changes in vital signs, haematology parameters, or blood chemistry or ECG abnormalities were observed. The estimated whole body effective dose of 68Ga-PSMA-11 (mean ± standard deviation) was 2.524 × 10-2 ± 2.546 × 10-3mSv/MBq. Time to maximum concentration (1.16 × 10-4 ± 1.3 × 10-5% ID/mL) in whole blood was 2.15 ± 0.33min. 68Ga-PSMA-11 has a favourable safety and tolerability profile in Japanese patients with primary, recurrent, or suspected recurrent prostate cancer, which is comparable to previous observations in other populations.

In vitro and preclinical systematic dose-effect studies of Auger electron- and beta particle-emitting radionuclides and external beam radiation for cancer treatment

Despite a rise in clinical use of radiopharmaceutical therapies, the biological effects of radionuclides and their relationship with absorbed radiation dose are poorly understood. Here, we set out to define this relationship for Auger electron-emitters [99mTc]TcO4 and [123I]I, and β--particle-emitter [188Re]ReO4. Studies were carried out using genetically-modified cells that permitted direct radionuclide comparisons. Triple-negative MDA-MB-231 breast cancer cells, expressing the human sodium/iodide symporter (hNIS) and green fluorescent protein (GFP; MDA-MB-231.hNIS-GFP) were used. In vitro radiotoxicity of [99mTc]TcO4, [123I]I and [188Re]ReO4 was determined using clonogenic assays. Radionuclide uptake, efflux, and subcellular location were used to calculate nuclear-absorbed doses using the Medical Internal Radiation Dose formalism. In vivo studies were performed using female NSG mice bearing orthotopic MDA-MB-231.hNIS-GFP tumors and compared to X-ray-treated (12.6-15 Gy) and untreated cohorts. Absorbed dose per unit activity in tumors and NIS-expressing organs were extrapolated to reference human adult models using OLINDA/EXM®. [99mTc]TcO4- and [123I]I reduced the survival fraction only in hNIS-expressing cells, whereas [188Re]ReO4 reduced survival fraction in hNIS-expressing and parental cells. [123I]I required 2.4-fold and 1.5-fold lower decays/cell to achieve 37% survival compared to [99mTc]TcO4- and [188Re]ReO4, respectively, following 72 hours incubation. Additionally, [99mTc]TcO4-, [123I]I and [188Re]ReO4 had superior cell killing effectiveness in vitro compared to X-rays. In vivo, X-ray led to a greater median survival compared to [188Re]ReO4 and [123I]I (54 days versus 45 and 43 days, respectively). Unlike the X-ray cohort, no metastases were visualized in the radionuclide-treated cohorts. Extrapolated human absorbed doses of [188Re]ReO4 to a 1 g tumor were 13.8-fold and 11.2-fold greater than for [123I]I in female and male models, respectively. This work reports reference dose-effect data using cell and tumor models for [99mTc]TcO4, [123I]I, and [188Re]ReO4, for the first time. We further demonstrate the tumor controlling effects of [123I]I, and [188Re]ReO4 in comparison to EBRT.

A study on out-of-field leakage of an accelerator-based neutron beam for boron neutron capture therapy.

More and more accelerator-based boron neutron capture therapy (AB-BNCT) facilities are under the construction or commissioning stage, and the neutron beam characteristic measurements at each facility will start soon. In addition to the in-field neutron beam properties, the leakage of neutron beam is also important, which is related to the side effects of the patient. In the Virtual Technical Meeting on Advances in Boron Neutron Capture Therapy held by International Atomic Energy Agency (IAEA) in July 2020, the issue of the out-of-field leakage in BNCT was addressed. Heron Neutron Medical Corporation has been working on the beam design for China Medical University Hsinchu Hospital AB-BNCT research center. To evaluate the out-of-field leakage, both beam profile analysis and whole-body dose calculation are performed. An Oak Ridge National Laboratory (ORNL) Medical Internal Radiation Dose (MIRD) mathematical phantom is used to calculate the whole-body dose. For the estimated irradiation time which is set to be the time required for 80% of tumor dose to reach 20 Gy-w, the relative biological effectiveness weighted dose of abdomen region is less than 40 mGy-w and the whole-body dose is 104 mSv. The beam profile calculational result shows that the neutron ambient dose equivalent at 15 cm from the field edge is 11 mSv/Gy-w and drops to 5 mSv/Gy-w at 26 cm from the field edge. The gamma ray ambient dose equivalent is less than 1 mSv/Gy-w starting from 10 cm from the field edge. Although the neutron out-of-field leakage of the beam design is higher than that of the initially proposed guideline by IAEA in 2020, the whole-body dose, however, is reasonably low. Both the whole-body dose evaluation and the beam profile analysis are useful in the beam design consideration. The whole-body dose calculation together with the beam profile analysis can also be helpful in reaching an acceptable recommendation for the out-of-field leakage for BNCT neutron beam, a job wished to be accomplished in the near future as proposed in the 2023 IAEA's report on Advances in Boron Neutron Capture Therapy.

Rapid predictive dosimetry for radioembolization.

Economics of today's busy clinical practice demand both time and cost-efficient methods of predictive dosimetry for liver radioembolisation. A rapid predictive schema adapted from the Medical Internal Radiation Dose (MIRD) method i.e., Partition Model, has been devised that can be completed within minutes. This rapid schema may guide institutions that do not have access to software capable of comprehensive auto-segmentation of lung, tumour and non-tumorous liver, or where rigorous artery-specific tomographic predictive dosimetry is unfeasible for the routine clinical workflow. This rapid schema is applicable to any beta-emitting radiomicrosphere, although absorbed dose-response thresholds will differ according to device. Sampling errors in lung, tumour and non-tumorous liver will compound and propagate throughout this schema. This rapid schema achieves efficiency in lieu of accuracy. The user must be mindful of potentially large sampling errors and assumes all responsibility. Any suspicion of significant error requires the user to revert back to standard-of-care methods.

A Review of Organ Dose Calculation Methods and Tools for Patients Undergoing Diagnostic Nuclear Medicine Procedures

Exponential growth has been observed in nuclear medicine procedures worldwide in the past decades. The considerable increase is attributed to the advance of positron emission tomography and single photon emission computed tomography, as well as the introduction of new radiopharmaceuticals. Although nuclear medicine procedures provide undisputable diagnostic and therapeutic benefits to patients, the substantial increase in radiation exposure to nuclear medicine patients raises concerns about potential adverse health effects and calls for the urgent need to monitor exposure levels. In the current article, model-based internal dosimetry methods were reviewed, focusing on Medical Internal Radiation Dose (MIRD) formalism, biokinetic data, human anatomy models (stylized, voxel, and hybrid computational human phantoms), and energy spectrum data of radionuclides. Key results from many articles on nuclear medicine dosimetry and comparisons of dosimetry quantities based on different types of human anatomy models were summarized. Key characteristics of seven model-based dose calculation tools were tabulated and discussed, including dose quantities, computational human phantoms used for dose calculations, decay data for radionuclides, biokinetic data, and user interface. Lastly, future research needs in nuclear medicine dosimetry were discussed. Model-based internal dosimetry methods were reviewed focusing on MIRD formalism, biokinetic data, human anatomy models, and energy spectrum data of radionuclides. Future research should focus on updating biokinetic data, revising energy transfer quantities for alimentary and gastrointestinal tracts, accounting for body size in nuclear medicine dosimetry, and recalculating dose coefficients based on the latest biokinetic and energy transfer data.

Dosimetry of [18F]TRACK, the first PET tracer for imaging of TrkB/C receptors in humans

BackgroundReduced expression or impaired signalling of tropomyosin receptor kinases (Trk receptors) are found in a vast spectrum of CNS disorders. [18F]TRACK is the first PET radioligand for TrkB/C with proven in vivo brain penetration and on-target specific signal. Here we report dosimetry data for [18F]TRACK in healthy humans. 6 healthy participants (age 22–61 y, 3 female) were scanned on a General Electric Discovery PET/CT 690 scanner. [18F]TRACK was synthesized with high molar activities (Am = 250 ± 75 GBq/µmol), and a dynamic series of 12 whole-body scans were acquired after injection of 129 to 147 MBq of the tracer. Images were reconstructed with standard corrections using the manufacturer’s OSEM algorithm. Tracer concentration time-activity curves (TACs) were obtained using CT-derived volumes-of-interest. Organ-specific doses and the total effective dose were estimated using the Committee on Medical Internal Radiation Dose equation for adults and tabulated Source tissue values (S values).ResultsAverage organ absorbed dose was highest for liver and gall bladder with 6.1E−2 (± 1.06E−2) mGy/MBq and 4.6 (± 1.18E−2) mGy/MBq, respectively. Total detriment weighted effective dose EDW was 1.63E−2 ± 1.68E−3 mSv/MBq. Organ-specific TACs indicated predominantly hepatic tracer elimination.ConclusionTotal and organ-specific effective doses for [18F]TRACK are low and the dosimetry profile is similar to other 18F-labelled radio tracers currently used in clinical settings.

A personalized Monte Carlo study of tumor and critical organ doses for Trans-arterial Radioembolization patients.

Trans-arterial Radioembolization (TARE) is an intra-arterial treatment method for liver malignancies. In this procedure, the therapeutic tumor dose is significant for predicting the treatment effectiveness while the dose absorbed in an organ at risk provides an understanding of its tolerance to radiation. This study proposes a Monte Carlo (MC) approach for determining absorbed organ doses for patients undergoing TARE treatment. The technique is based on the use of a voxel-based partial body model generated for each patient from his/her anatomical image data to represent the critical body structures more realistically. These structures are first segmented from image slices to create an image block which is then incorporated into a radiation transport package (MCNP6.2) to perform MC simulations. When used along with the parameters specific to a patient's treatment, such as lung-shunt factor (LSF), tumor-to-normal liver ratio (TLR), fractional uptakes, and administered activity, this approach allowed more accurate simulation of radiation interactions and hence provided absorbed doses specific to a TARE patient. The MC method also calculated the absorbed doses in organs or tissues that were close to target tissues for which the MIRD formalism makes no predictions. MIRD calculations were found to overestimate the absorbed doses by as much as 11% in lungs, 5% in liver, and 20% in tumor volumes. This raises concerns about the treatment's efficacy when estimating the correct activity to be administered to a patient. When each patient simulation was repeated with a 90Y source spectrum to reflect the distribution of varying beta energies, the liver and the lungs were observed to receive relatively lower doses than those obtained with monoenergetic beta particles. Thus, it can be stated that the approach adopted in this study offers a more precise model of the patient's critical tissues and serves as a personalized dosimetric tool for TARE treatment planning.

BNCT dosimetry for brain cancer based on the nuclear research reactor NUR and the modified MIRD phantom by introducing the eye lens: Simulation study

The Glioblastoma Multiforme (GBM) is one of the most aggressive type of the brain cancer, as well as resistant to treatments combining surgery and concomitant sessions of radiotherapy, based on X-rays, and chemotherapy. The Boron Neutron Capture Therapy (BNCT) has been elected as a radical solution to this type of cancer. This present work focuses on the dosimetric study of the GBM using the therapeutic BNCT beam based on the NUR nuclear research reactor of the Draria Nuclear Research Center (CRND) that has been already optimized in a previous research. In this context, the simulations have been done using the Monte Carlo Code (MCNP5) and the Medical Internal Radiation Dose (MIRD) phantom with an assumption of a brain tumor of 1.5 cm radius situated at different depth ranging from 4 to 10 cm and centered with respect to the neutron beam. In addition, the calculation of the absorbed dose by the tumor, the healthy brain tissue, the skin around head and neck (skin), the pharynx and the thyroid was performed for the different locations of the tumor using the ICRU 46 to determine the elemental composition and the flux to Kerma conversion coefficients for each tissue. However, the recent epidemiological studies show that cataract can occur at a dose lower than 200 cGy, as it can be a stochastic effect and not a deterministic one. To this end, a calculation of the absorbed dose by the lens of the eye and more precisely by its most radiosensitive part (germinative zone) was developed by modifying the MIRD phantom to model the entire eye based on the equations of later studies. The obtained results show that the absorbed dose rate calculated for the tumor increases each time its depth rises, which is not the case for the dose rate calculated for the healthy organs where it remains approximately constant. The absorbed dose by the germinative zone varies from (68.37 ± 0.89) cGy for tumor depth of 4 cm to (733.92 ± 15.96) cGy for a tumor situated at 10 cm of depth.

Presentation of Organ Dose and Effective Dose Conversion Factors in Dual-Energy Computed Tomography: A Monte Carlo Simulation Study.

The same conversion factors (k-factors) of Single CT (SECT) are applied to estimate the Effective Dose (ED) in Dual Energy Computed Tomography (DECT). However, k-factors for different organs need independently validating for DECT, due to the different conditions in DECT. This study aimed to calculate organ dose and k-factors in different imaging protocols (liver, chest, cardiac, and abdomen) for male and female phantoms. This Monte Carlo Simulation study used Monte Carlo N-Particle (MCNP) code for modeling a Siemens Somatom Definition Flash dual-source CT scanner. The organ dose, dose length product, and k-factors were calculated for the Medical Internal Radiation Dose (MIRD) of male and female phantoms. For the male phantom, the k-factors for the liver, chest, cardiac, and abdomen-pelvis imaging protocols are equal to 0.020, 0.012, 0.016, and 0.014 mSv.mGy-1cm-1, respectively. For the female phantom, the corresponding values are equal to 0.026, 0.023, 0.036, and 0.018, respectively. These values for DECT are different from those corresponding values for SECT, especially for the female phantom. The calculated k-factors for DECT can be used as reference values for the estimation of ED in DECT.

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.

Comparison of organ dose from chest radiography with varying beam quality and constant exposure index.

The optimum X-ray tube voltage for chest radiographic examinations remains unclear; hence, the tube voltage varies between medical facilities. An exposure index (EI) was proposed to standardize the parameters for radiographic examinations. However, even if identical EI values are used to examine the same person, organ doses may vary due to differences in tube voltages. In this study, the variation in organ doses between different beam qualities under identical EI values for chest radiographic examinations was investigated using Monte Carlo simulations. A focused anti-scatter grid as well as standard and larger physique-type medical internal radiation dose (MIRD) phantoms were studied under tube voltages of 90, 100, 110, and 120 kVp. The organ doses in the MIRD phantom increased as the X-ray tube voltage decreased, even with identical EI values. The absorbed doses in the lungs of standard and large-sized MIRD phantoms at 90 kVp were 23% and 35% higher than those at 120 kVp, respectively. The doses to organs other than the lung at 90 kVp were also higher than those at 120 kVp. From the perspective of reducing radiation doses, a tube voltage of 120 kVp is considered better for chest examinations compared with a tube voltage of 90 kVp under identical EI values.

Development of a stand-alone precalculated Monte Carlo code to calculate the dose by alpha and beta emitters from the Ra-224 decay chain.

Recent developments in alpha and beta emitting radionuclide therapy highlight the importance of developing efficient methods for patient-specific dosimetry. Traditional tabulated methods such as Medical Internal Radiation Dose (MIRD) estimate the dose at the organ level while more recent numerical methods based on Monte Carlo (MC) simulations are able to calculate dose at the voxel level. A precalculated MC (PMC) approach was developed in this work as an alternative to time-consuming fully simulated MC. Once the spatial distribution of alpha and beta emitters is determined using imaging and/or numerical methods, the PMC code can be used to achieve an accurate voxelized 3D distribution of the deposited energy without relying on full MC calculations. To implement the PMC method to calculate energy deposited by alpha and beta particles emitted from the Ra-224 decay chain. The GEANT4 (version 10.7) MC toolkit was used to generate databases of precalculated tracks to be integrated in the PMC code as well as to benchmark its output. In this regard, energy spectra of alpha and beta particles emitted by the Ra-224 decay chain were generated using GAMOS (version 6.2.0) and imported into GEANT4 macro files. Either alpha or beta emitting sources were defined at the center of a homogeneous phantom filled with various materials such as soft tissue, bone, and lung where particles were emitted either mono-directionally (for database generation) or isotropically (for benchmarking). Two heterogeneous phantoms were used to demonstrate PMC code compatibility with boundary crossing events. Each precalculated database was generated step-by-step by storing particle track information from GEANT4 simulations followed by its integration in a PMC code developed in MATLAB. For a user-defined number of histories, one of the tracks in a given database was selected randomly and rotated randomly to reflect an isotropic emission. Afterward, deposited energy was divided between voxels based on step length in each voxel using a ray-tracing approach. The radial distribution of deposited energy was benchmarked against fully simulated MC calculations using GEANT4. The effect of the GEANT4 parameter StepMax on the accuracy and speed of the code was also investigated. In the case of alpha decay, primary alpha particles show the highest contribution (>99%) in deposited energy compared to their secondary particles. In most cases, protons act as the main secondary particles in the deposition of energy. However, for a lung phantom, using a range cutoff parameter of 10µm on primary alpha particles yields a higher contribution of secondary electrons than protons. Differences between deposited energy calculated by PMC and fully simulated MC are within 2% for all alpha and beta emitters in homogeneous and heterogeneous phantoms. Additionally, statistical uncertainties are less than 1% for voxels with doses higher than 5% of the maximum dose. Moreover, optimization of the parameter StepMax is necessary to achieve the best tradeoff between code accuracy and speed. The PMC code shows good performance for dose calculations deposited by alpha and beta emitters. As a stand-alone algorithm, it is suitable to be integrated into clinical treatment planning systems.

Radiation dosimetry and pharmacokinetics of the tau PET tracer florzolotau (18F) in healthy Japanese subjects

ObjectiveAbnormal aggregation of tau in the brain is a major contributing factor in various neurodegenerative diseases. Florzolotau (18F) (florzolotau, APN-1607, PM-PBB3) has been shown to be a probe for tau fibrils in an animal model and patients with Alzheimer’s disease and those with non-Alzheimer’s disease tauopathies. The objective of this study is to evaluate the safety, pharmacokinetics, and radiation dose following a single intravenous administration of florzolotau in healthy Japanese subjects.MethodsThree healthy male Japanese subjects aged between 20 and 64 were enrolled in this study. Subjects were determined to be eligible based on the screening assessments at the study site. Subjects received a single intravenous dose of 195.0 ± 0.5 MBq of florzolotau and underwent the whole-body PET scan 10 times in total to calculate absorbed doses to major organs/tissues and effective dose. Radioactivities in whole blood and urine were also measured for pharmacokinetic evaluation. Absorbed doses to major organs/tissues and effective dose were estimated using the medical internal radiation dose (MIRD) method. Vital signs, electrocardiography (ECG), and blood tests were done for safety evaluation.ResultsThe intravenous injection of florzolotau was well tolerated. There were no adverse events or clinically detectable pharmacologic effects related to the tracer in any subjects. No significant changes in vital signs and ECG were observed. The highest mean initial uptake at 15 min after injection was in the liver (29.0 ± 4.0%ID), intestine (4.69 ± 1.65%ID), and brain (2.13 ± 0.18%ID). The highest absorbed dose was 508 μGy/MBq of the gallbladder wall, followed by the liver of 79.4 μGy/MBq, the pancreas of 42.5 μGy/MBq, and the upper large intestine of 34.2 μGy/MBq. The effective dose was calculated as 19.7 μSv/MBq according to the tissue weighting factor reported by ICRP-103.ConclusionFlorzolotau intravenous injection was well tolerated in healthy male Japanese subjects. The effective dose was determined as 3.61 mSv when 185 MBq florzolotau was given.

Evaluating bone marrow dosimetry with the addition of bone marrow structures to the medical internal radiation dose phantom

AbstractBackgroundReliable estimates of radiation dose to bone marrow are critical to understanding the risk of radiation‐induced cancers. Although the medical internal radiation dose phantom is routinely used for dose estimation, bone marrow is not defined in the phantom. Consequently, methods of indirectly estimating bone marrow dose have been implemented based on dose to surrogate volumes or average dose to soft tissue.MethodsIn this study, new bone marrow structures were implemented and evaluated to the medical internal radiation dose phantom in Geant4, offering improved fidelity. The dose equivalent to the bone marrow was calculated across medical, occupational, and space radiation exposure scenarios, and compared with results using prior indirect estimation methods.ConclusionOur results show that bone marrow dose may be overestimated by up to a factor of three when using the traditional methods when compared with the improved fidelity medical internal radiation dose method, specifically at clinical x‐ray energies.

Theranostic cobalt-55/58m for neurotensin receptor-mediated radiotherapy in vivo: A pilot study with dosimetry

Neurotensin receptor 1 (NTSR1) can stimulate tumor proliferation through neurotensin (NTS) activation and are overexpressed by a variety of cancers. The high binding affinity of NTS/NTSR1 makes radiolabeled NTS derivatives interesting for cancer diagnosis and staging. Internalization of NTS/NTSR1 also suggests therapeutic application with high LET alpha particles and low energy electrons. We investigated the therapeutic efficacy of [58mCo]Co-NOTA-NT-20.3 in vivo using murine models xenografted with NTSR1-positive HT29 human colorectal adenocarcinoma cells, and utilized [55Co]Co-NOTA-NT-20.3 for dosimetry. MethodsTargeting properties and cytotoxicity of [55/58mCo]Co-NOTA-NT-20.3 were assessed with HT29 cells. Female nude mice were xenografted with HT29 tumors and administered [55Co or 58mCo]Co-NOTA-NT-20.3 to evaluate pharmacokinetics or for therapy, respectively. Dosimetry calculations followed the Medical Internal Radiation Dose (MIRD) formalism and human absorbed dose rate per unit activity were obtained from OpenDose. The pilot therapy study consisted of two groups (each N = 3) receiving 110 ± 15 MBq and 26 ± 6 MBq [58mCo]Co-NOTA-NT-20.3 one week after tumor inoculation, and control (N = 3). Tumor sizes and masses were measured twice a week after therapy. Complete blood count and kidney histology were also performed to assess toxicity. ResultsHPLC measured radiochemical purity of [55,58mCo]Co-NOTA-NT-20.3 > 99 %. Labeled compounds retained NTS targeting properties. [58mCo]Co-NOTA-NT-20.3 exhibited cytotoxicity for HT29 cells and was >15× more potent than [58mCo]CoCl2. Xenografted tumors responded modestly to administered doses, but mice showed no signs of radiotoxicity. Absorbed dose to tumor and kidney with 110 MBq [58mCo]Co-NOTA-NT-20.3 were 0.6 Gy and 0.8 Gy, respectively, and other organs received less than half of the absorbed dose to tumor. Off-target radiation dose from cobalt-58g was small but reduces the therapeutic window. ConclusionThe enhanced in vitro cytotoxicity and high tumor-to-background led us to investigate the therapeutic efficacy of [58mCo]Co-NOTA-NT-20.3 in vivo. Although we were unable to induce tumor response commensurate with [177Lu]Lu-NT127 (NLys-Lys-Pro-Tyr-Tle-Leu) studies involving similar time-integrated activity, the absence of observed toxicity may constitute an opportunity for targeting vectors with improved uptake and/or retention to avoid the aftereffects of other high-LET radioactive emissions. Future studies with higher uptake, activity and/or multiple dosing regimens are warranted. The theranostic approach employed in this work was crucial for dosimetry analysis.

Radiation pneumonitis following Yttrium-90 radioembolization: A Korean multicenter study.

To report the incidence of radiation pneumonitis after radioembolization. In this retrospective study, from May 2009 to July 2021, 782 consecutive patients underwent radioembolization in two institutes. Medical internal radiation dose dosimetry and partition dosimetry were used for glass and resin Yttrium-90-labeled microspheres (90Y-microspheres), respectively. Medical records and radiological findings were retrospectively evaluated with emphasis on the symptomatic radiation pneumonitis. Of the 732 patients with lung shunt study and follow-up, 13 (1.8%) had symptomatic radiation pneumonitis and six patients died due to radiation pneumonitis. Of the 721 patients whose lung doses were calculated, 10 patients who were treated with glass (n = 5) and resin (n = 5) 90Y-microspheres had radiation pneumonitis. No significant statistical difference between glass and resin 90Y-microspheres (p = 0.304) was noted in terms of radiation pneumonitis incidence. Among the patients with radiation pneumonitis, all five patients treated with glass 90Y-microspheres had estimated lung doses > 29 Gy, whereas five patients treated with resin 90Y-microspheres had relatively wide range of lung dose reaching much lower value (13.21Gy). The present study suggests that radiation pneumonitis after radioembolization may occur even though the manufacturer's instructions are followed.

Ongoing Updates to NNDC Web Services

The National Nuclear Data Center (NNDC) provides nuclear data through 40 websites, many of which are being updated to make data easier to find. The most visible change is a site-wide navigation menu that links to the NNDC’s most-visited databases. The Evaluated Nuclear Structure Data File (ENSDF) front page now also has a chart interface for visual exploration of its data. NuDat 3.0 provides a faster, smoother, and feature-rich interactive Chart of Nuclides. The Medical Internal Radiation Dose (MIRD) website has been updated with improved response times and decay diagrams. More recently, the NNDC has focused on standardizing its web development process. Through the use of Git version control and Gradle Build Tool, the NNDC intends to improve existing websites and guide the creation of new ones.

Concepts and methods for the dosimetry of radioembolisation of the liver with Y-90-loaded microspheres

This article aims at presenting in a didactic way, dosimetry concepts and methods that are relevant for radio-embolization of the liver with 90Y-microspheres. The application of the medical internal radiation dose formalism to radio-embolization is introduced. This formalism enables a simplified dosimetry, where the absorbed dose in a given tissue depends on only its mass and initial activity. This is applied in the single-compartment method, partition model, for the liver, tumour and lung dosimetry, and multi-compartment method, allowing identification of multiple tumours. Voxel-based dosimetry approaches are also discussed. This allows taking into account the non-uniform uptake within a compartment, which translates into a non-uniform dose distribution, represented as a dose–volume histogram. For this purpose, dose–kernel convolution allows propagating the energy deposition around voxel-sources in a computationally efficient manner. Alternatively, local-energy deposition is preferable when the spatial resolution is comparable or larger than the beta-particle path. Statistical tools may be relevant in establishing dose–effect relationships in a given population. These include tools such as the logistic regression or receiver operator characteristic analysis. Examples are given for illustration purpose. Moreover, tumour control probability modelling can be assessed through the linear-quadratic model of Lea and Catcheside and its counterpart, the normal-tissue complication probability model of Lyman, which is suitable to the parallel structure of the liver. The selectivity of microsphere administration allows tissue sparing, which can be considered with the concept of equivalent uniform dose, for which examples are also given. The implication of microscopic deposition of microspheres is also illustrated through a liver toxicity model, even though it is not clinically validated. Finally, we propose a reflection around the concept of therapeutic index (TI), which could help tailor treatment planning by determining the treatment safety through the evaluation of TI based on treatment-specific parameters.

Comparison of PMT-based TF64 and SiPM-based Vereos PET/CT systems for 90 Y imaging and dosimetry optimization: A quantitative study.

Selective internal radiotherapy based on transarterial radioembolization (TARE) with yttrium-90 (90 Y) microspheres is an established treatment for primary or metastatic liver disease. The objective of this work is to optimize the dosimetry of patients treated with 90 Y TARE, using positron emission tomography (PET) images. The NEMA 2012 PET phantom was filled with nearly 3.9GBq of 90 Y activity and acquired at days 0, 3, 5, 7, and 9 on a classic time-of-flight PET/computed tomography (CT) scanner (Philips TF64) and on a silicon photomultiplier (SiPM)-based PET/CT scanner (Philips Vereos). Acquisitions were carried on following the guidelines proposed in a previously published multicentric trial and images were reconstructed by varying and combining the available parameters. Comparisons were performed to identify the best set(s) of parameters leading to the most accurate 90 Y-PET image(s), in terms of activity distribution. Then, for both scanners, the best images were analyzed with Simplicit90 Y, a personalized dosimetry software using multicompartmental Medical Internal Radiation Dose model. The comparison between measured and true doses allowed to identify the image granting the most consistent dose estimations and, therefore, to designate the set of parameters to be applied on patients' data for the reconstruction of optimized clinical images. Posttreatment dosimetry of four patients was then realized with Simplicit90 Y using optimized imaging datasets. Based on activity distribution comparisons and dose estimations over phantom and patients data, the SiPM-based PET/CT system appeared more suitable than the photomultiplier tube-based TF64 for 90 Y-PET imaging. With the SiPM-based PET/CT system, reconstructed images with a 2-mm voxel size combined with the application of the point spread function correction led to the most accurate results for quantitative 90 Y measures. For the SiPM-based PET/CT scanner, an optimized set of reconstruction parameters has been identified and applied on patients' data in order to generate the most accurate image to be used for an improved personalized 90 Y-PET dosimetry, ensuring a reliable evaluation of the delivered doses.

EFFECTIVE DOSE TO ADULT PATIENTS UNDERGOING RENAL SCANS WITH 99MTC (DMSA, DTPA, EC AND MAG3).

Renal scintigraphy plays an important role in the diagnosis of various kidney disorders. This procedure can be performed with different radiopharmaceuticals. The patients undergoing renal scintigraphy receive a radiation dose that should be assessed. In this study, the effective dose of patients due to renal scintigraphy with 99mTc (DMSA, DTPA, EC and MAG3) was calculated for the International Commission on Radiological Protection (ICRP) 110 adult reference phantoms using the Medical Internal Radiation Dose method and the ICRP 60 and 103 tissue weighting factors. The results show that the highest effective dose per unit activity administered is due to 99mTc DMSA. On average, the effective dose per unit activity administered of 99mTc DMSA is almost twice the effective dose per unit activity administered of other radiopharmaceuticals. The effective doses per unit activity administered calculated using the ICRP 103 tissue weighting factors are always lower than that calculated using the ICRP 60 tissue weighting factors and the ICRP 128 data.