Internal Radiation Dose Research Articles

Neoadjuvant Yttrium-90 Transarterial Radioembolization with Resin Microspheres Prescribed Using the Medical Internal Radiation Dose Model for Intrahepatic Cholangiocarcinoma

PurposeTo evaluate outcomes of patients with intrahepatic cholangiocarcinoma (iCCA) undergoing neoadjuvant yttrium-90 (90Y) transarterial radioembolization (TARE) with resin microspheres prescribed using the Medical Internal Radiation Dose (MIRD) model. Materials and MethodsThis retrospective institutional review board–approved study included 37 patients with iCCA treated with 90Y-TARE from October 2015 to September 2020. The primary outcome was overall survival (OS) from 90Y-TARE. The secondary outcomes were progression-free survival (PFS), Response Evaluation Criteria In Solid Tumors 1.1 imaging response, and downstaging to resection. Patients with tumor proximity to the middle hepatic vein (<1 cm) and/or insufficient future liver remnant were treated with neoadjuvant intent (n = 21). Patients were censored at the time of surgery or at the last follow-up for the Kaplan-Meier survival analysis. ResultsFor 31 patients (69 years; interquartile range, 64–74 years; 20 men [65%]) included in the study, the first-line therapy was 90Y-TARE for 23 (74%) patients. Imaging assessment at 6 months showed a disease control rate of 86%. The median PFS was 5.4 months (95% confidence interval [CI], 3–not reached). The PFS was higher after first-line 90Y-TARE (7.4 months [95% CI, 5.3–not reached]) than that after subsequent 90Y-TARE (2.7 months [95% CI, 2–not reached]) (P = .007). The median OS was 22 months (95% CI, 7.3–not reached). The 1- and 2-year OS rates were 60% (95% CI, 41%–86%) and 40% (95% CI, 19.5%–81%). In patients treated with neoadjuvant intent, 11 of 21 patients (52%) underwent resections. The resection margins were R0 and R1 in 8 (73%) and 3 (27%) of 11 patients, respectively. On histological review in 10 patients, necrosis of ≥90% tumor was achieved in 7 of 10 patients (70%). ConclusionsFirst-line 90Y-TARE prescribed using the MIRD model as neoadjuvant therapy for iCCA results in good survival outcome and R0 resection for unresectable patients.

Radiation effects on residents caused by comprehensive utilization of slag in building materials

Objective This study intends to analyze the radioactivity level of slag and its building materials, estimate the effective dose of the residents caused by the comprehensive utilization of slag in building materials, and provide basis for rational utilization of slag. Methods Data of slag production and its comprehensive utilization in building materials, radioactive level investigation and detection of slag and building materials in different periods and regions were collected from 2005 to 2016. Using the room model of concrete structure and the model of calculating equilibrium equivalent radon concentration with the specific activity of 226Ra in indoor building materials, the external and internal radiation doses of residents caused by comprehensive utilization of slag for building materials were estimated respectively, and then the annual additional effective dose and collective dose of residents were estimated. Results The contents of 226Ra and 232Th in the slag were relatively high, while the contents of 40K were relatively low. The radionuclide content in slag cement decreased significantly. The annual additional effective doses of slag used in concrete and cement mixture relative to ordinary cement were 40 mSv · a –1 and 0.20 mSv · a –1 respectively. The annual additional collective dose and 50-year collective dose caused by slag used in building materials were 3.87 × 10 3~1.84 × 10 4 man · Sv and 1.94 × 10 5~9.20 × 10 5 man · Sv respectively. Conclusion The comprehensive utilization of slag in building materials increases the additional effective dose of residents, so the use of some slag in building materials needs to be paid more attention to. In addition, the influence of comprehensive utilization of slag in building materials, such as the way, the amount of slag and the amount of building materials, on the annual additional effective dose of residents caused should also be paid attention to. 摘要： 目的 本研宄拟分析矿渣及其建材的放射性水平, 估算矿渣综合利用于建材所致居民的有效剂量, 为合理利用 矿渣提供依据。 方法 收集 2005—2016 年矿渣产生量和综合利用于建材量的资料, 以及不同时期、不同地区的矿渣 及其建材放射性水平调查与检测的资料等。采用混凝土结构房间模型、室内建材中 226Ra 比活度推算平衡当量氡浓度 的模式分别估算矿渣综合利用于建材所致居民的外照射和内照射剂量, 进而估算居民年附加有效剂量和集体剂量。 结果 矿渣中 226Ra 和 232Th 的含量相对较高, 40K 含量较低, 矿渣水泥中的放射性核素含量较矿渣中明显降低; 矿渣用 于混凝土和水泥混合材料相对普通水泥住房导致公众附加年有效剂量分别为0.40 mSv · a –1 和 0.20 mSv · a –1, 矿渣用 于建材所致公众年附加集体剂量和 50 年集体剂量分别为 3.87 × 10 3~1.84 × 10 4 人· Sv 和 1.94 × 10 5~9.20 × 10 5 人· Sv。 结论 矿渣综合利用于建材使居民附加有效剂量有所增加, 利用某些矿渣生产建材需要引起一定的重 视;此外, 矿渣综合利用于建材的方式、利用量、建材使用量等对其所致居民年附加有效剂量的影响也需关注。

Internal Radiation Dose Evaluation for an Unruptured Post Release Tristructural Istropic Fuel Particle for Advanced and Micro-reactor Applications: Erratum.

Internal Radiation Dose Evaluation for an Unruptured Post Release Tristructural Istropic Fuel Particle for Advanced and Micro-reactor Applications: Erratum.

Patient dosimetry of 177Lu-PSMA I&T in metastatic prostate cancer treatment: the experience in Thailand.

This study aimed to determine the radiation dosimetry for 177Lu-PSMA imaging and therapy (I&T) in Thai patients who were treated for metastatic prostate cancer. Whole-body planar images acquired at immediately, 4 and 24h after 177Lu-PSMA I&T injection (range 4.44-8.51GBq) were collected from 12 treatment cycles of 8 prostate cancer patients. Region of interests (ROIs) were manually contoured on the whole-body, liver, spleen, urinary bladder, lacrimal glands, parotid, and submandibular glands to determine time-integrated activity (TIA) in source organs and fitted time-activity curves using mono-exponential extrapolation. The S values calculated utilizing non-uniform rational B-splines (NURBS) computational phantoms were extracted from the OLINDA/EXM v. 2.0 to calculate the absorbed dose coefficient in target organs according to the Medical Internal Radiation Dose (MIRD) scheme. The absorbed doses to bone marrow were estimated using the planar two-compartment image-based method by separating the high-uptake and low-uptake compartment. The spherical model was used to calculate the lacrimal gland absorbed doses. Mean absorbed dose coefficients to the kidneys, bone marrow, liver, urinary bladder, spleen, lacrimal glands, parotid, and submandibular glands were 0.81 ± 0.24, 0.02 ± 0.01, 0.13 ± 0.10, 0.27 ± 0.25, 0.16 ± 0.07, 3.62 ± 1.78, 0.21 ± 0.14, and 0.09 ± 0.07Gy/GBq, respectively. Dose constraints for the kidneys (23Gy) and bone marrow (2Gy) were not reached in any patients. The absorbed dose in lacrimal glands calculated by the NURBS computational phantoms was slightly lower than the calculation based on the Cristy-Eckerman computational phantoms using OLINDA/EXM v. 1.0 by 6.37 ± 0.14%. Dosimetry results in this study suggested that 177Lu-PSMA I&T treatment with higher activities and more cycles is possible without the risk of damaging normal organs in prostate cancer patients.

Internal dosimetry in F-18 FDG PET examinations based on long-time-measured organ activities using total-body PET/CT: does it make any difference from a short-time measurement?

PurposeA 2-m axial field-of-view, total-body PET/CT scanner (uEXPLORER) has been recently developed to provide total-body coverage and ultra-high sensitivity, which together, enables opportunities for in vivo time-activity curve (TAC) measurement of all investigated organs simultaneously with high temporal resolution. This study aims at quantifying the cumulated activity and patient dose of 2-[F-18]fluoro-2-deoxy-D-glucose (F-18 FDG ) imaging by using delayed time-activity curves (TACs), measured out to 8-h post-injection, for different organs so that the comparison between quantifying approaches using short-time method (up to 75 min post-injection) or long-time method (up to 8 h post-injection) could be performed.MethodsOrgan TACs of 10 healthy volunteers were collected using total-body PET/CT in 4 periods after the intravenous injection of F-18 FDG. The 8-h post-injection TACs of 6 source organs were fitted using a spline method (based on Origin (version 8.1)). To compare with cumulated activity estimated from spline-fitted curves, the cumulated activity estimated from multi-exponential curve was also calculated. Exponential curve was fitted with shorter series of data consistent with clinical procedure and previous dosimetry works. An 8-h dynamic bladder wall dose model considering 2 voiding were employed to illustrate the differences in bladder wall dose caused by the different measurement durations. Organ absorbed doses were further estimated using Medical Internal Radiation Dose (MIRD) method and voxel phantoms.ResultsA short-time measurement could lead to significant bias in estimated cumulated activity for liver compared with long-time-measured spline fitted method, and the differences of cumulated activity were 18.38% on average. For the myocardium, the estimated cumulated activity difference was not statistically significant due to large variation in metabolism among individuals. The average residence time differences of brain, heart, kidney, liver, and lungs were 8.38%, 15.13%, 25.02%, 23.94%, and 16.50% between short-time and long-time methods. Regarding effective dose, the maximum differences of residence time between long-time-measured spline fitted curve and short-time-measured multi-exponential fitted curve was 9.93%. When using spline method, the bladder revealed the most difference in the effective dose among all the investigated organs with a bias up to 21.18%. The bladder wall dose calculated using a long-time dynamic model was 13.79% larger than the two-voiding dynamic model, and at least 50.17% lower than previous studies based on fixed bladder content volume.ConclusionsLong-time measurement of multi-organ TACs with high temporal resolution enabled by a total-body PET/CT demonstrated that the clinical procedure with 20 min PET scan at 1 h after injection could be used for retrospective dosimetry analysis in most organs. As the bladder content contributed the most to the effective dose, a long-time dynamic model was recommended for the bladder wall dose estimation.

Prediction of Absorbed Dose to Normal Organs with Endocrine Tumors for I-131 by use of 99mTC Single Photon Emission Computed Tomography/Computed Tomography and Geant4 Application for Tomographic Emission Simulation.

Introduction:This study aimed to predict the dose absorbed by normal organs with neuroendocrine tumors for 131I using single photon emission computed tomography/computed tomography (SPECT/CT) images and Geant4 application for tomographic emission (GATE) simulation.Materials and Methods:Four to 5 whole-body planar scan series, along with one SPECT/CT image, were taken from four patients following 99mTc-hynic-Tyr3-octreotide radiotracer injection. After image quantification, the residence time of each organ was calculated using the image analysis and the activity time curves. The energy deposit and dose conversion (S-value) were extracted from the GATE simulation for the target organs of each patient. Using the residence times and S-values, the mean absorbed dose for the target organs of each patient was calculated and compared with the data obtained from the standard method.Results:Very close agreement was obtained between the S-value of the self–organ irradiation. The mean percentage difference between the two methods (i.e. GATE and Medical Internal Radiation Dose [MIRD]) was 1.8%, while a weak agreement was observed for cross-organ irradiation. The percentage difference between the total absorbed doses by the organs was 2%. The percentage difference between the absorbed doses obtained for tumors and three considered normal organs estimated by the GATE method was slightly higher than the MIRD method (about 11% on average for tumors).Conclusion:Regardless of the small difference between the obtained results for the organs and absorbed doses of the tumors in the present study, patient-specific dosimetry by the GATE methods is useful and essential for therapeutic radionuclides such as 131I due to high cross-dose effects, especially for young adult patients, to ensure the radiation safety and increase the effectiveness of the treatment.

Monte Carlo determination of dose coefficients at different developmental stages of zebrafish (Danio rerio) in experimental condition

The release of liquid effluent of nuclear power into aquatic system increases with the rapid development of nuclear facilities in coastal and inland regions. Aquatic model animals are very important for the study of the radiation hazards to non-human biota in water environment and its extrapolation of dose-effect relationship to human models. However, the study of the radiation dose rate calculation model of the aquatic animal zebrafish is still on the homogeneous isotropic model used for the protection of the environment. A series of zebrafish models (including adults, larvae and embryos, named zebrafish-family: ZF-family) with multiple internal organs are established in this study to investigate the mechanism of radiation damage effect in order to protect non-human species. The internal and external dose coefficients (DCs) of the whole body, heart and gonads of zebrafishes are calculated in water environment with the combination of the real experimental culture condition, using Monte Carlo application package GATE (Geant4 Application for Emission Tomography) and eight nuclides, i.e., 3H, 14C, 90Sr, 60Co, 110mAg, 134Cs, 137Cs, 131I, which are commonly found in the liquid effluent of nuclear power plants, as the source items, The results show that the level of nuclide γ energy determines the external DCs (DCext), and 90Sr plays the most important role in internal DCs (DCint). The comparison between the external DCs of the heart and gonad and that of the whole body shows that DCs (DCext) of heart and gonad for females are 80% and 43% lower than that of whole body, respectively, while for males, the DCs (DCext) of heart is 44% lower than that of the whole body, and DCs (DCext) of gonad is slightly higher than that of the whole body for most nuclides (up to 25%).The dose of internal radiation makes greater contribution than that of external radiation to pure beta emitter (3H, 14C, 90Sr). This internal DCs of ZF-family model with complex internal structure turns out to demonstrate more sensitive DCs change trend and higher calculation values compared with the internal DCs of the simple ellipsoid model. In this model, the photon emitter with strong penetrating power has higher internal DCs, while the low-energy pure beta nuclide does not alter much. In conclusion, it is vital to carry out refined systematic modeling for model organisms, and the determination of DCs of model organs can promote the evaluation of the radiation effects on non-human species.

Biological impacts on the lungs in rats internally exposed to radioactive 56MnO2 particle

To understand the radiation effects of the atomic bombing of Hiroshima and Nagasaki among the survivors, radiation from neutron-induced radioisotopes in soil should be considered in addition to the initial radiation directly received from the bombs. 56Mn, which emits both β particles and γ-rays, is one of the dominant radioisotopes created in soil by neutrons from the bomb. Thus we investigated the biological effects of internal exposure to 56MnO2 particle in the lung of male Wistar rats comparing to the effects of external 60Co-γ irradiation. Absorbed doses of internal irradiation of lungs were between 25 and 65 mGy in 56MnO2-exposed animals, while the whole body doses were between 41 and 100 mGy. Animals were examined on days 3 and 61 after the exposure. There were no remarkable pathological changes related to 56MnO2 particle exposure. However, mRNA and protein expressions of aquaporin 5 increased significantly in the lung tissue on day 3 postexposure in 56MnO2 groups (by 1.6 and 2.9 times, respectively, in the highest dose group). Smad7 mRNA expression was also significantly elevated by 30% in the highest dose group of 56MnO2. Our data demonstrated that internal exposure to 56MnO2 induced significant biological responses including gene expression changes in the lungs, while external 60Co-γ irradiation of 2 Gy did not show any changes.

Prospective evaluation of organ-specific dose and lesional doses following therapeutic [177Lu]Lu-EDTMP administration in patients with multiple skeletal metastases and its correlation with clinical hematological toxicity.

In patients with multiple skeletal metastases, accurate estimation of absorbed doses to radiosensitive bone marrow in bone-directed systemic radionuclide therapies (RNT) is critically important from clinical dose determination standpoint. The primary aim of the present study was to estimate the radiation absorbed doses of therapeutic [177Lu]Lu-EDTMP to bone marrow by two methods viz. Medical Internal Radiation Dose (MIRD) schema and using OLINDA software and correlate with hematological toxicity. A total of 15 patients diagnosed to have multiple painful skeletal metastases being treated with [177Lu]Lu-EDTMP for palliation of pain, were enrolled for this prospective study. For all patients, urine was collected immediately after infusion of [177Lu]Lu-EDTMP up to 24 h post-administration and cumulative activity excreted from body via urine was calculated. For dosimetry, patients underwent post-administration whole-body scintigraphy at five-time points: 0.5 (pre-void), 2, 24, 48 and 120 h (post-void). From the time-activity curves generated by drawing regions of interest (ROIs) on the images, number of disintegrations was determined. Absorbed doses for organs and bone lesions were calculated using OLINDA 2.2.0 software. For bone marrow dose estimates, in addition to OLINDA 2.2.0 software, MIRD schema was also adopted. Hematological profile was monitored in all patients during the treatment and post-treatment follow-up (estimating complete blood counts, every 15 d for 3 months after therapy). The mean ± standard deviation activity of [177Lu]Lu-EDTMP administered per patient per cycle was 2.08 ± 0.45 GBq. The results demonstrated higher uptake of [177Lu]Lu-EDTMP in bone metastases compared to normal bones. Within 2 and 24 h of administration of [177Lu]Lu-EDTMP, [177Lu]Lu activity excreted from the body was 24 ± 9% and 39 ± 14%, respectively. The mean absorbed organ doses (mean ± SD) in Gy/GBq were as follows: osteogenic cells 3.15 ± 1.85, bone marrow 0.57 ± 0.31, kidneys 0.08 ± 0.05, urinary bladder 0.32 ± 0.04, and bone lesions 2.91 ± 1.88. Strong correlation was found between (a) MIRD schema and OLINDA 2.2.0 software method for estimation of bone marrow doses (r = 0.96; P = <0.0001) and (b) Bone marrow absorbed dose and hematological toxicity (r = 0.81, P = 0.0027). Radiation absorbed doses to the bone marrow and skeletal metastatic lesions, following therapeutic [177Lu]Lu-EDTMP were estimated using a convenient and non-invasive quantitative imaging method. The estimated bone marrow absorbed dose, either by MIRD schema or the OLINDA 2.2.0 software method, demonstrated strong correlation. Strong correlation was also observed between bone marrow absorbed dose and hematological toxicity.

Activity concentrations of 238U, 232Th, 226Ra 137Cs and 40K radionuclides in honey samples from Romania. Lifetime cancer risk estimated

Romania is on the first place in Europe with an annual production of honey of 30.9 thousand tons in 2018, while the European Union is on the second place in the world with a total of 283 thousand tons. The most part of the honey production from Romania is exported.The aim of this study was to estimate internal radiation doses and lifetime cancer risk from honey ingestion sampled from several locations on the Romanian territory.The activity concentration of 137Cs and 238U is found to be below the detection limit of the method. The activity of 226Ra ranged from lower than the minimum detectable value (<1.08 Bq·kg−1) to 7.35 ± 1.73 Bq·kg−1. For 232Th and 40K the average recorded values were 1.51 ± 0.74 and 24.08 ± 4.37 Bq·kg−1, respectively.The annual effective dose due to the intake of 232Th, 226Ra, 40K has been assessed ranging from 0.08 to 0.47 (μSv·y−1), 0.35 to 1.65 (μSv·y−1) and 0.04 to 0.33 (μSv·y−1), respectively. 26Ra gives the largest contribution (56–82%) to the total annual effective dose due to ingestion received by consumption of honey, in the samples where it was identified. In the other samples (where 226Ra was below the detection limit), the largest contribution to annual effective dose is due to 232Th (45–85%).The values obtained from the lifetime cancer risk calculation range in the domain from 2.85·10−7 to 1.78·10−6 for 232Th, 1.33·10−6 to 6.22·10−6 for 226Ra and 1.38·10−7 to 1.25·10−6 for 40K. For 137Cs, only one value was recorded 2.20·10−8.The 232Th, 226Ra, 40K and 137Cs activities measured in the present work were comparable with completion of others activated values of honey samples around Europe.

Evaluation Internal Radiation Dose of Pediatric Patients during Medicine Tests Using Monte Carlo Simulation

Evaluation Internal Radiation Dose of Pediatric Patients during Medicine Tests Using Monte Carlo Simulation

Radiation monitoring of drinking water in the vicinity of the Beloyarsk NPP

The article provides a radiation-hygienic assessment of the current state of drinking water supply sources for the population in the observation area of the the Beloyarsk NPP and the Institute of Nuclear Materials. We determined the content of natural (234U,238U,226Ra,228Ra,210Po,222Rn,210Pb,228Th,230Th,232Th) and technogenic (3H,14C,60Co,90Sr,134Cs,137Cs,238Pu,239,240Pu,241Am) radionuclides in drinking water of tap water, water boreholes and water wells in test settlements located at different distances and directions from radiation hazardous facilities. Results of monitoring of water sources in 2012–2013 and 2019 showed the radiation safety of drinking water in the vicinity of the Beloyarsk NPP according to several criteria. Thus, the maximum levels of the gross specific alpha-activity of radionuclides in water samples were 3.9 times lower than the control level (0.2 Bq/kg), the gross specific beta-activity was 5.7 times lower than the control level (1 Bq/ kg). Over the entire observation period, none of the drinking water samples exceeded the control levels both for individual radionuclides and for the sum of the ratios of specific activities to control levels. The content of natural and artificial radionuclides in drinking water near the Beloyarsk NPP decreases in the following order: water wells &gt; water boreholes &gt; tap water. For the past 20 years, there was a decrease in tritium specific activity in drinking water of the Beloyarsk NPP region by 20–35%, depending on the source of water supply. It was noted that the launch of the BN-800 reactor also did not lead to an increase in the content of other artificial radionuclides (90Sr,137Cs) in groundwater. The average annual effective dose of internal exposure of the population due to drinking water consumption in the vicinity of the Beloyarsk NPP is 0.05 mSv, according to conservative estimates – 0.07 mSv, which is below the radiation safety threshold (0.1 mSv/a) recommended by the WHO. Natural radionuclides play the primary role in the formation of the annual average effective dose for internal irradiation (98.9%) due to drinking water consumption on the considered territories.210Po makes the largest contribution to the dose from natural radioisotopes – 43%, somewhat less is made by210Pb – 25%. The third place in the dose formation from natural radionuclides belongs to234U (8%),228Ra (7%),226Ra (6%) and230Th (6%). The contribution of other natural radioisotopes in the formation of the internal radiation dose from drinking water consumption does not exceed 2-3%. The contribution of technogenic radionuclides to the annual average effective dose from the consumption of drinking water is negligible (about 1%). Of the technogenic components,90Sr (60%),3H (20%), and241Am (12%) play the most significant role in the formation of the internal exposure dose.

Assessment of conservatism of assumptions on the population exposure conditions applied for establishing the permissible atmospheric release levels

The annual permissible atmospheric release levels of radionuclides are established with the condition that the population annual dose should not exceed the dose quota. Two conservative assumptions on the population exposure are applied in handbooks for calculation of the permissible release levels: a person can reside in any point outside the sanitary protection zone of nuclear facility; all exposure pathways, including food consumption, are determined by the place of residence. Calculation of the annual dose for adults living near the Smolensk, Rostov and Bilibino NPP was performed for three scenarios to evaluate the degree of conservatism of these assumptions. Scenario 1 assumed permanent residence in a critical point of the area with production of all food in this point and consumption only local food by the population. Scenario 2 assumed that a certain part of food is imported. In scenario 3 dose assessment was made for a permanent resident of the closest settlement near NPP, taking into consideration realistic characteristics of the exposure conditions. Atmospheric release levels of radionuclides assessed during the radiation and technical surveys of the NPP’s emission sources in 2017 – 2019 were used as the initial data. It is shown that the assumptions on the conditions of population exposure significantly affect the conservatism of the annual dose assessment. The calculated annual dose for the most conservative scenario is higher than for realistic scenario by 6,3 times (for the Smolensk NPP), 8,4 times (Rostov NPP), 883 times (Bilibino NPP). The highest degree of conservatism was shown for the internal radiation dose estimates. The results of calculation of the annual dose from NPP’s emissions strongly depend on the method for dose assessment from releases of 3H and 14C. Use of the regional values of consumption of locally produced food allows reasonable reducing the conservatism of the annual dose estimates from the NPP’s permanent radionuclide releases.

Estimating of internal radiation doses due to food consumption and its reduction applying the food regulation after the Fukushima nuclear accident using national food-monitoring data

Estimating of internal radiation doses due to food consumption and its reduction applying the food regulation after the Fukushima nuclear accident using national food-monitoring data

Description of the methodology for dosimetric quantification in treatments with 177Lu-DOTATATE

Implementation of dosimetry calculations in the daily practice of Nuclear Medicine Departments is, at this time, a controversial issue, partly due to the lack of a standardized methodology that is accepted by all interested parties (patients, nuclear medicine physicians and medical physicists). However, since the publication of RD 601/2019 there is a legal obligation to implement it, despite the fact that it is a complex and high resource consumption procedure. The aim of this article is to review the theoretical bases of in vivo dosimetry in treatments with 177Lu-DOTATATE. The exposed methodology is the one proposed by the MIRD Committee (Medical Internal Radiation Dose) of the SNMMI (Society of Nuclear Medicine & Molecular Imaging). According to this method, the absorbed dose is obtained as the product of 2 factors: the time-integrated activity of the radiopharmaceutical present in a source region and a geometrical factor S. This approach, which a priori seems simple, in practice requires several SPECT/CT acquisitions, several measurements of the whole body activity and taking several blood samples, as well as hours of image processing and computation. The systematic implementation of these calculations, in all the patients we treat, will allow us to obtain homogeneous data to correlate the absorbed doses in the lesions with the biological effect of the treatment. The final purpose of the dosimetry calculations is to be able to maximize the therapeutic effect in the lesions, controlling the radiotoxicity in the organs at risk.

Descripción de la metodología para la cuantificación dosimétrica en tratamientos con 177Lu-DOTATATE

Implementation of dosimetry calculations in the daily practice of Nuclear Medicine Departments is, at this time, a controversial issue, partly due to the lack of a standardized methodology that is accepted by all interested parties (patients, nuclear medicine physicians and medical physicists). However, since the publication of RD 601/2019 there is a legal obligation to implement it, despite the fact that it is a complex and high resource consumption procedure. The aim of this article is to review the theoretical bases of in vivo dosimetry in treatments with 177Lu-DOTATATE. The exposed methodology is the one proposed by the MIRD Committee (Medical Internal Radiation Dose) of the SNMMI (Society of Nuclear Medicine & Molecular Imaging). According to this method, the absorbed dose is obtained as the product of 2factors: the time-integrated activity of the radiopharmaceutical present in a source region and a geometrical factor S. This approach, which a priori seems simple, in practice requires several SPECT/CT acquisitions, several measurements of the whole body activity and taking several blood samples, as well as hours of image processing and computation. The systematic implementation of these calculations, in all the patients we treat, will allow us to obtain homogeneous data to correlate the absorbed doses in the lesions with the biological effect of the treatment. The final purpose of the dosimetry calculations is to be able to maximize the therapeutic effect in the lesions, controlling the radiotoxicity in the organs at risk.

68Ga]Ga-iPSMA-Lys3-Bombesin: Biokinetics, dosimetry and first patient PET/CT imaging

BackgroundThe prostate-specific membrane antigen (PSMA) and the gastrin-releasing peptide receptor (GRPR) are overexpressed in prostate cancer (PCa). In preclinical studies, the iPSMA-Lys3-Bombesin (iPSMA-BN) heterodimeric ligand has shown a suitable affinity for PSMA and GRPR. This research aimed to assess the biokinetics and radiation dosimetry of [68Ga]Ga-iPSMA-BN in four healthy volunteers based on biodistribution data obtained from whole-body PET/CT studies, as well as to visualize the [68Ga]Ga-iPSMA-BN tumor uptake in a patient with PCa. MethodsPET/CT images acquired at 5 min, 0.5, 1, and 2 h after radiotracer administration (124.5 ± 2.1 MBq) were corrected for attenuation, scattering, dead-time, and decay. The activity in the segmented volumes of interest (VOIs) in each source organ at different times was adjusted to mono- and bi-exponential biokinetic models (A(t)VOI), from which the total disintegrations (N) were calculated to assess the internal radiation doses by using the OLINDA V1.1 code. ResultsImages from the patient showed an evident uptake by the metastasis (SUVmax of 4.7) and by the organs expressing GRPR (pancreas) and PSMA (salivary glands). The average effective dose was 2.70 ± 0.05 mSv, which was like those known for most of the 68Ga studies, making [68Ga]Ga-iPSMA-BN a promising dual-target PET imaging radiotracer for PCa. Conclusions[68Ga]Ga-iPSMA-BN, capable of detecting both PSMA and GRPR with suitable biokinetics and dosimetric patterns, could be a potential complementary diagnostic tool for the improvement of prostate cancer PET imaging.

Influence of dosimetry method on bone lesion absorbed dose estimates in PSMA therapy: application to mCRPC patients receiving Lu-177-PSMA-I&T

BackgroundPatients with metastatic, castration-resistant prostate cancer (mCRPC) present with an increased tumor burden in the skeleton. For these patients, Lutetium-177 (Lu-177) radioligand therapy targeting the prostate-specific membrane antigen (PSMA) has gained increasing interest with promising outcome data. Patient-individualized dosimetry enables improvement of therapy success with the aim of minimizing absorbed dose to organs at risk while maximizing absorbed dose to tumors. Different dosimetric approaches with varying complexity and accuracy exist for this purpose. The Medical Internal Radiation Dose (MIRD) formalism applied to tumors assumes a homogeneous activity distribution in a sphere with unit density for derivation of tumor S values (TSV). Voxel S value (VSV) approaches can account for heterogeneous activities but are simulated for a specific tissue. Full patient-individual Monte Carlo (MC) absorbed dose simulation addresses both, heterogeneous activity and density distributions. Subsequent CT-based density weighting has the potential to overcome the assumption of homogeneous density in the MIRD formalism with TSV and VSV methods, which could be a major limitation for the application in bone metastases with heterogeneous density. The aim of this investigation is a comparison of these methods for bone lesion dosimetry in mCRPC patients receiving Lu-177-PSMA therapy.ResultsIn total, 289 bone lesions in 15 mCRPC patients were analyzed. Percentage difference (PD) of average absorbed dose per lesion compared to MC, averaged over all lesions, was + 14 ± 10% (min: − 21%; max: + 56%) for TSVs. With lesion-individual density weighting using Hounsfield Unit (HU)-to-density conversion on the patient’s CT image, PD was reduced to − 8 ± 1% (min: − 10%; max: − 3%). PD on a voxel level for three-dimensional (3D) voxel-wise dosimetry methods, averaged per lesion, revealed large PDs of + 18 ± 11% (min: − 27%; max: + 58%) for a soft tissue VSV approach compared to MC; after voxel-wise density correction, this was reduced to − 5 ± 1% (min: − 12%; max: − 2%).ConclusionPatient-individual MC absorbed dose simulation is capable to account for heterogeneous densities in bone lesions. Since the computational effort prevents its routine clinical application, TSV or VSV dosimetry approaches are used. This study showed the necessity of lesion-individual density weighting for TSV or VSV in Lu-177-PSMA therapy bone lesion dosimetry.

Radiation dose rate to Japanese cedar and plants collected from Okuma, Fukushima Prefecture.

After the 2011 Fukushima Dai-ichi Nuclear Power Station (FDNPS) accident, wild populations of animals and plants living in the evacuation zone received additional ionizing radiation of both internal and external radiation doses. Morphological abnormalities of pine and fir trees near the FDNPS were reported. In order to evaluate dose-effect relationships, it is necessary to quantify the radiation doses to trees and plants. In this study, the internal and external dose rates to Japanese cedar and plants collected at three sites in Okuma, approximately 4km southwest of FDNPS were estimated applying the ERICA Assessment Tool. The activity concentrations of 134Cs and 137Cs in soils, cedar trunks, and plants were determined. The total dose rates to cedar ranged from 2.2±1.2 to 6.1±2.2 μGy h-1. These rates were within the derived consideration reference levels (DCRLs) reported by ICRP 108 as 4-40 μGy h-1 for pine trees. The highest estimate for plants was 7.1±2.7 μGy h-1, much smaller than the DCRLs reported for grasses and herbs (40-400 μGy h-1). On average, the internal radiation dose rates to cedars at the two sites accounted for 5% and 29% of the external dose rates, respectively, while the value in another site was only 0.4% for cedar. This was attributed to differences in the crown area between the three sites. The trunk diameter of cedars shows a positive correlation with the ratio of internal to external radiation dose rates. It indicates that the total dose rate to cedars is easily estimated with the soil radiocaesium inventory and trunk diameter. The internal radiation dose rate to the plant varied depending on the plant species. This variation was considerably large in plants due to the presence of two species, including Solidago altissima and Artemisia indica var. maximowiczii.

Previous Malignancy as a Risk Factor for the Second Solid Cancer in a Cohort of Nuclear Workers

This paper describes the results of epidemiological analysis of a cohort of nuclear workers hired at the main facilities of “Mayak” Production Association located in the city of Ozyorsk in Southern Urals of the Russian Federation. Previous malignancy as a risk factor for second cancer in a cohort of 22,373 workers occupationally exposed to ionizing radiation has been analyzed in a retrospective study with more than 60 years of follow-up. Information on main radiation and non-radiation risk factors (attained age, sex, tobacco smoking), as well as the dose of exposure to occupational ionizing radiation has been obtained for the analysis using the data from cancer register as well as other main population registries created in Epidemiological Laboratory of Southern Urals Biophysics Institute. Poisson’s regression realized in the “Amfit” module of “Epicure” statistical package has been applied for risk analysis. Excess relative risk per 1 Gy of absorbed dose of external gamma radiation and internal alpha radiation has been calculated using linear model. Among the 2,471 cancer cases accumulated in the study cohort to the end of follow-up 6.4% of second cancer cases have been diagnosed among workers occupationally exposed to protracted external gamma- and internal alpha radiation. The relative risk of second cancer (except for non-melanoma skin cancer) among nuclear workers with previously diagnosed cancer was about 4 times higher after a decade compared with those cancer patients who had single cancer only. The results showed that previous malignancy along with main non-radiation factors is statistically significant carcinogenic risk factor among nuclear workers exposed to protracted occupational radiation. Doi: 10.28991/SciMedJ-2021-0301-2 Full Text: PDF