Nuclear Medicine Dose Research Articles

Establishment and Application of a Radiation Dose Rate Model for Nuclear Medicine Examinees.

Traditional methods for determining radiation dose in nuclear medicine include the Monte Carlo method, the discrete ordinate method, and the point kernel integration method. This study presents a new mathematical model for predicting the radiation dose rate in the vicinity of nuclear medicine patients. A new algorithm was created by combining the physical model of "cylinder superposition" of the human body with integral analysis to assess the radiation dose rate in the vicinity of nuclear medicine patients. The model accurately predicted radiation dose rates within distances of 0.1-3.0 m, with a deviation of less than 11% compared to observed rates. The model demonstrated greater accuracy at shorter distances from the radiation source, with a deviation of only 1.55% from observed values at 0.1 m. The model proposed in this study effectively represents the spatial and temporal distribution of the radiation field around nuclear medicine patients and demonstrates good agreement with actual measurements. This model has the potential to serve as a radiation dose rate alert system in hospital environments.

Development of the occupational exposure during the production and application of radiopharmaceuticals in Germany

An increasing number of radiopharmaceuticals and proteins are available for diagnosing and treating various diseases. The demand for existing and newly developed pharmaceutical radionuclides and proteins is steadily increasing. The radiation exposure levels of workers in the radiopharmaceutical industry and nuclear medicine field are closely monitored, specifically their effective dose and equivalent dose, leading to the question, of whether the dawn of radiopharmaceuticals affects the occupational exposure level. This development is analyzed and evaluated with data from the German National Dose Register. Data shows that the effective dose in the work categories production and distribution of radioisotopes as well as nuclear medicine slightly decreased from 1997 to 2021. Over the same period, the hand equivalent dose in nuclear medicine increases steadily, with no discernible trend in production and distribution of radioisotopes. Over the past few decades, intentional efforts and measures have been taken to ensure radiation protection. Instruments for monitoring and dose reduction must be continuously applied. Given the low effective dose, the focus in future shall be on dose reduction following the as low as reasonably achievable principle. The development of the hand equivalent dose should be carefully observed in the upcoming years.

Evaluating the feasibility of Compton imaging towards improved spatial resolution and reduced patient dose in nuclear medicine

Evaluating the feasibility of Compton imaging towards improved spatial resolution and reduced patient dose in nuclear medicine

Evaluation of institutional whole-body and extremity occupational radiation doses in nuclear medicine.

This study evaluated nuclear medicine occupational radiation doses at Sultan Qaboos University Hospital, a 700-bed tertiary care teaching hospital in Oman. Personal effective whole-body doses, Hp(10), and extremity doses, Hp(0.07), were collected for 19 medical radiation workers over a 7-year period (2015-2021). Personal doses for four professional groups were measured using calibrated thermo-luminescence dosemeters ((LiF:Mg,Ti) TLD-100). The average, median and maximum cumulative doses were compared against the annual whole-body and extremity dose limits (20mSv and 500mSv y-1, respectively) and local dose investigation level (DIL; 6mSv y-1). Personal whole-body doses (average:median:maximum) for technologists, medical physicists, nuclear medicine physicians and nurses were 1.8:1.1:7.8, 0.3:0.3:0.4, 0.1:0.1:0.2 and 0.1:0.1:0.2mSv, respectively. Personal extremity doses for left and right hand (average and maximum doses) follow similar trends. Average annual effective whole-body and extremity doses were well below the recommended annual dose limits. The findings suggest lowering local DIL for all staff except for technologists.

Determination of a reliable assessment for occupational eye lens dose in nuclear medicine.

The most recent statement published by the International Commission on Radiological Protection describes a reduction in the maximum allowable occupational eye lens dose from 150 to 20 mSv/year (averaged over 5‐year periods). Exposing the eye lens to radiation is a concern for nuclear medicine staff who handle radionuclide tracers with various levels of photon energy. This study aimed to define the optimal dosimeter and means of measuring the amount of exposure to which the eye lens is exposed during a routine nuclear medicine practice. A RANDO human phantom attached to Glass Badge and Luminess Badge for body or neck, DOSIRIS and VISION for eyes, and nanoDot for body, neck, and eyes was exposed to 99mTc, 123I, and 18F radionuclides. Sealed syringe sources of each radionuclide were positioned 30 cm from the abdomen of the phantom. Estimated exposure based on measurement conditions (i.e., air kerma rate constants, conversion coefficient, distance, activity, and exposure time) was compared measured dose equivalent of each dosimeter. Differences in body, neck, and eye lens dosimeters were statistically analyzed. The 10‐mm dose equivalent significantly differed between the Glass Badge and Luminess Badge for the neck, but these were almost equivalent at the body. The 0.07‐mm dose equivalent for the nanoDot dosimeters was greatly overestimated compared to the estimated exposure of 99mTc and 123I radionuclides. Measured dose equivalents of exposure significantly differed between the body and eye lens dosimeters with respect to 18F. Although accurately measuring radiation exposure to the eye lenses of nuclear medicine staff is conventionally monitored using dosimeters worn on the chest or abdomen, eye lens dosimeters that provide a 3‐mm dose equivalent near the eye would be a more reliable means of assessing radiation doses in the mixed radiation environment of nuclear medicine.

Individual dosimetry system for targeted alpha therapy based on PHITS coupled with microdosimetric kinetic model

BackgroundAn individual dosimetry system is essential for the evaluation of precise doses in nuclear medicine. The purpose of this study was to develop a system for calculating not only absorbed doses but also EQDX(α/β) from the PET-CT images of patients for targeted alpha therapy (TAT), considering the dose dependence of the relative biological effectiveness, the dose-rate effect, and the dose heterogeneity.MethodsA general-purpose Monte Carlo particle transport code PHITS was employed as the dose calculation engine in the system, while the microdosimetric kinetic model was used for converting the absorbed dose to EQDX(α/β). PHITS input files for describing the geometry and source distribution of a patient are automatically created from PET-CT images, using newly developed modules of the radiotherapy package based on PHITS (RT-PHITS). We examined the performance of the system by calculating several organ doses using the PET-CT images of four healthy volunteers after injecting 18F-NKO-035.ResultsThe deposition energy map obtained from our system seems to be a blurred image of the corresponding PET data because annihilation γ-rays deposit their energies rather far from the source location. The calculated organ doses agree with the corresponding data obtained from OLINDA 2.0 within 20%, indicating the reliability of our developed system. Test calculations by replacing the labeled radionuclide from 18F to 211At suggest that large dose heterogeneity in a target volume is expected in TAT, resulting in a significant decrease of EQDX(α/β) for higher-activity injection.ConclusionsAs an extension of RT-PHITS, an individual dosimetry system for nuclear medicine was developed based on PHITS coupled with the microdosimetric kinetic model. It enables us to predict the therapeutic and side effects of TAT based on the clinical data largely available from conventional external radiotherapy.

Design of monitoring system for radioactivity and radiation dose in nuclear medicine

In this paper, an intelligent radioactivity and radiation dose monitoring system is designed, which can be used in the field of nuclear medicine. The system can dynamically monitor the body surface radioactivity and radiation dose rate of patients receiving 131I radioactive drug treatment, and can identify the personnel who exceed the radiation threshold and send out sound and light alarm. Compared with other monitoring equipment, this system has higher accuracy and can guide the clinical treatment of nuclear medicine scientifically.

Radiation dose in nuclear medicine: the hybrid imaging.

Hybrid imaging procedures such as single-photon emission computed tomography/computed tomography (SPECT/CT) and positron emission tomography/computed tomography (PET/CT) showed a rapid diffusion in recent years because of their high sensitivity, specificity, and accuracy, due to a more accurate localization and definition of scintigraphic findings. However, hybrid systems inevitably lead to an increase in patient radiation exposure because of the added CT component. Effective doses due to the radiopharmaceuticals can be estimated by multiplying the administered activities by the effective dose coefficients, while for the CT component the dose-length product can be multiplied by a conversion coefficient k. However, the effective dose value is subject to a high degree of uncertainty and must be interpreted as a broad, generic estimate of biologic risk. Although the effective dose can be used to estimate and compare the risk of radiation exposure across multiple imaging techniques, clinicians should be aware that it represents a generic evaluation of the risk derived from a given procedure to a generic model of the human body. It cannot be applied to a single individual and should not be used for epidemiologic studies or the estimation of population risks due to the inherent uncertainties and oversimplifications involved. Practical ways to reduce radiation dose to patients eligible for hybrid imaging involve adjustments to both the planning phase and throughout the execution of the study. These methods include individual justification of radiation exposure, radiopharmaceutical choice, adherence to diagnostic reference levels (DLR), patient hydration and bladder voiding, adoption of new technical devices (sensitive detectors or collimators) with new reconstruction algorithms, and implementation of appropriate CT protocols and exposure parameters.

319. Effective Dose Assessment on an Hybrid Equipment: Results and challenges to integrate with MDCT

Purpose The aim of this study was to measure, monitor and report effective dose in nuclear medicine and to assess current challenges in integrating these measures with CT. Methods In 2014 we connected a CT-PET to a dose tracking software and started to collect dose data from CT modality; in late 2015 the software was upgraded to a version which includes a nuclear medicine module for dose tracking. Data from the PET part of the scanner, in terms of administered activity and effective dose, were also collected. Results Data from 12000 CT doses and 3700 PET doses were collected, mostly from: Whole Body (WB) 18F-FDG (63.2%), WB 11C-Choline (12.8%), WB 68Ga-DOTATOC (5.5%), Head 11C-Methionine for a total of 18.4% of the whole activity of the scanner. CT mean DLP (mGy∗cm) data values were (range between brackets): WB 18F-FDG 1486.3 (76.1-7700.9), WB 11C-Choline 2451.9 (290.9-5023.1), WB 68Ga-DOTATOC 1183.8 (41.0-3728.4) and Head 11C-Methionine 766.7 (52.3-2326.9). Mean effective doses (mSv) were: WB 18F-FDG 15.3 ± 4.7, WB 11C-Choline 26.5 ± 5.6, WB 68Ga 13.4 ± 3.7 and Head 11C-Methionine 1.2 ± 0.7. Mean administered activities (MBq) and effective doses (mSv) for radiopharmaceuticals used in PET were: WB 18F-FDG 335.9-6.4, Head 11C-Methionine 294.6-2.5, WB 68Ga-DOTATOC 168.3-3.5 and WB 11C-Choline 322.2–1.5. The mean total effective doses (mSv) due to both modality were: WB 18F-FDG 21.7 ± 4.9, Head 11C-Methionine 3.7 ± 0.8, WB 11 C-Choline 27.9 ± 5.6 and WB 68Ga-DOTATOC 16.9 ± 2.8. Conclusions The implementation of a dose tracking system to nuclear medicine is of great value for accurate and regular recording, reporting and analysis of patient’s effective doses. It can help to improve the evaluation of radiation exposures in the clinical practice.

OA144] Measuring, monitoring and reporting effective dose on hybrid equipment: Results and challanges to integrate with MDCT in a large population study

Purpose The aim of this study was to measure, monitor and report the effective dose in nuclear medicine on s large population and to integrate these measurements with CT dose data. Methods and materials A dose tracking software (GE Dosewatch™) with a NM module was connected to a CT-PET and collected DLP and effective dose from CT modality and administered activity and effective from PET modality. The mean DLP and the mean effective doses with the ranges was assessed for CT. The mean administered activities and mean effective doses for the different radiopharmaceuticals used in PET were evaluated. Examinations were performed with a GE Discovery 690 which has a 64-detector CT scanner; PET scan is performed in stop and go imaging mode. Results Doses data from 11000 CT and 2700 PET were collected, mostly from: Whole Body (WB) 18F-FDG (63.2%), WB 11C-Choline (12.8%), Head 11C-Methionine (18.4%), WB 68Ga-DOTATOC (5.5%). CT mean DLP (mGy*cm) data values were (range between brackets): WB 18F-FDG 1486.3 (76.1-7700.9), WB 11C-Choline 2451.9 (290.9-5023.1), WB 68Ga-DOTATOC 1184.2 (41.0-3728.4) and Head 11C-Methionine 766.7 (52.3-2326.9). Mean effective doses due to CT modality (mSv) were: WB 18F-FDG 15.3 ± 4.7, WB 11C-Choline 26.5 ± 5.6, WB 68Ga-DOTATOC 13.4 ± 2.7 and Head 11C-Methionine 1.7 ± 2.9. Mean administered activity (MBq) and effective doses (mSv) for radiopharmaceuticals used in PET were: WB 18F-FDG 335.9-6.4, Head 11C-Methionine 294.4.6-2.5, WB 68Ga-DOTATOC 168.3-3.5 and WB 11C-Choline 322.2-1.5. The mean total effective doses (mSv) due to both modality were: WB 18F-FDG 21.7 ± 4.9, Head 11C-Methionine 4.3 ± 2.9, WB 11C-Choline 27.9 ± 5.6 and WB 68Ga-DOTATOC 16.9 ± 2.8. Conclusions The implementation of a dose tracking system in nuclear medicine is of great value for accurate and regular recording, reporting and analysis of patient’s effective doses. It can help to improve and optimize the evaluation of radiation exposures in the clinical practice.

A U.S. Multicenter Study of Recorded Occupational Radiation Badge Doses in Nuclear Medicine.

Purpose To summarize occupational badge doses recorded for a sample of U.S. nuclear medicine technologists. Materials and Methods Nine large U.S. medical institutions identified 208 former and current nuclear medicine technologists certified after 1979 and linked these individuals to historic badge dose records maintained by a commercial dosimetry company (Landauer), yielding a total of 2618 annual dose records. The distributions of annual and cumulative occupational doses were described by using summary statistics. Results Between 1992 and 2015, the median annual personal dose equivalent per nuclear medicine technologist was 2.18 mSv (interquartile range [IQR], 1.25-3.47 mSv; mean, 2.69 mSv). Median annual personal dose equivalents remained relatively constant over this period (range, 1.40-3.30 mSv), while maximum values generally increased over time (from 8.00 mSv in 1992 to 13.9 mSv in 2015). The median cumulative personal dose equivalent was 32.9 mSv (IQR, 18.1-65.5 mSv; mean, 51.4 mSv) for 45 technologists who had complete information and remained employed through 2015. Conclusion Occupational radiation doses were well below the established occupational limits and were consistent with those observed for nuclear medicine technologists worldwide and were greater than those observed for nuclear and general medical workers in the United States These results should be informative for radiation monitoring and safety efforts in nuclear medicine departments. © RSNA, 2018 Online supplemental material is available for this article.

Abstract ID: 225 Development of adjustable model breast using mammographic information for Monte Carlo simulation in nuclear medicine — Impact on dosimetry

For accurate estimation of the breast dose in nuclear medicine examinations, the development of an adjustable sizes model breast for Monte Carlo simulation is important. The development of the model breast for nuclear medicine studies is difficult because the size of the breast must be known. The purpose of this study is to develop the adjustable sizes model breast using mammographic information for Monte Carlo simulation in nuclear medicine and to investigate its impact on dosimetry using the SimDOSE simulation code [1] . This study constructed two sizes of ORNL phantoms with two radionuclides: 18F and 99mTc. The dimension of a breast phantom, a semi-oblate spheroid model, was established from the collected breast parameters of a cohort of patients in mammography [2] to provide a model breast allowing for changing sizes. Breast phantoms with various breast sizes, shapes, compositions, and the glandular distribution were created. The preliminary results showed the ratios of S-values of the breast ranged from 0.94 to 1.01 and the difference caused by breast shapes was insignificant. The organ dose of left and right breast may be considered separately due to in the dose difference between left and right breast. For different glandular distributions, the differences in S-values of the gland self-absorbed dose is about 5–8%. We conclude that the breast size has significant impacts and the glandular distribution has moderate impacts on the S-values of the breast, while the breast shape and compositions cause minor changes in the S-value. Therefore, this study provided the new method to assess the breast dose in Nuclear Medicine.

Feasibility of reducing differences in estimated doses in nuclear medicine between a patient-specific and a reference phantom

The feasibility of reducing the differences between patient-specific internal doses and doses estimated using reference phantoms was evaluated. Relatively simple adjustments to a polygon-surface ICRP adult male reference phantom were applied to fit selected individual dimensions using the software Rhinoceros®4.0. We tested this approach on two patient-specific phantoms: the biggest and the smallest phantoms from the Helmholtz Zentrum München library. These phantoms have unrelated anatomy and large differences in body-mass-index. Three models approximating each patient’s anatomy were considered: the voxel and the polygon-surface ICRP adult male reference phantoms and the adjusted polygon-surface reference phantom. The Specific Absorbed Fractions (SAFs) for internal photon and electron sources were calculated with the Monte Carlo code EGSnrc. Employing the time-integrated activity coefficients of a radiopharmaceutical (S)-4-(3-18F-fluoropropyl)-l-glutamic acid and the calculated SAFs, organ absorbed-dose coefficients were computed following the formalism promulgated by the Committee on Medical Internal Radiation Dose. We compared the absorbed-dose coefficients between each patient-specific phantom and other models considered with emphasis on the cross-fire component. The corresponding differences for most organs were notably lower for the adjusted reference models compared to the case when reference models were employed. Overall, the proposed approach provided reliable dose estimates for both tested patient-specific models despite the pronounced differences in their anatomy. To capture the full range of inter-individual anatomic variability more patient-specific phantoms are required. The results of this test study suggest a feasibility of estimating patient-specific doses within a relative uncertainty of 25% or less using adjusted reference models, when only simple phantom scaling is applied.

Internal dosimetry with the Monte Carlo code GATE: validation using the ICRP/ICRU female reference computational model

The purpose of this work was to validate GATE-based clinical scale absorbed dose calculations in nuclear medicine dosimetry.GATE (version 6.2) and MCNPX (version 2.7.a) were used to derive dosimetric parameters (absorbed fractions, specific absorbed fractions and S-values) for the reference female computational model proposed by the International Commission on Radiological Protection in ICRP report 110.Monoenergetic photons and electrons (from 50 keV to 2 MeV) and four isotopes currently used in nuclear medicine (fluorine-18, lutetium-177, iodine-131 and yttrium-90) were investigated.Absorbed fractions, specific absorbed fractions and S-values were generated with GATE and MCNPX for 12 regions of interest in the ICRP 110 female computational model, thereby leading to 144 source/target pair configurations. Relative differences between GATE and MCNPX obtained in specific configurations (self-irradiation or cross-irradiation) are presented. Relative differences in absorbed fractions, specific absorbed fractions or S-values are below 10%, and in most cases less than 5%. Dosimetric results generated with GATE for the 12 volumes of interest are available as supplemental data.GATE can be safely used for radiopharmaceutical dosimetry at the clinical scale. This makes GATE a viable option for Monte Carlo modelling of both imaging and absorbed dose in nuclear medicine.

Awareness of ionizing radiation exposure among junior doctors and senior medical students in radiological investigations

The awareness and knowledge of ionizing radiation exposure in radiological investigations among junior doctors and medical students were studied. The participants were year four to year six senior medical students enrolled at University of Dammam and interns in King Fahad University Hospital. The survey consisted of 22 questions designed online using the software ‘QuestionPro’ licensed to the University of Dammam. 100 hard copies were also distributed manually and collected. A total of 221 (88.5%) questionnaires were completed. 213 participants viewed, 151 started and 128 (84.7%) completed online. 93% of the distributed samples were completed. Overall knowledge was poor; 44% and 19% of the respondents thought incorrectly that MRI and ultrasound emit ionizing radiation, respectively. Respondents (92%; n = 203) underestimated the dose of abdominal spiral computed tomography (CT) and 4% thought no ionizing radiation involved in CT. 59% of respondents underestimated the radiation doses in nuclear medicine; bone scan 87%, PET/CT scan 67%, thyroid isotope scan 45% and PET scan 36%. 47% of the subjects had attended formal lectures, tutorials or workshops on radiation protection while 53% (n = 119) had not. For future education the majority stated they would prefer tutorials or workshops (42.3%) or problem-based learning/case studies (32.4%), while web-based modules would be their last choice (8.1%).

Evaluation of a deterministic grid-based Boltzmann solver (GBBS) for voxel-level absorbed dose calculations in nuclear medicine

To evaluate the 3D Grid-based Boltzmann Solver (GBBS) code ATTILA® for coupled electron and photon transport in the nuclear medicine energy regime for electron (beta, Auger and internal conversion electrons) and photon (gamma, x-ray) sources. Codes rewritten based on ATTILA are used clinically for both high-energy photon teletherapy and 192Ir sealed source brachytherapy; little information exists for using the GBBS to calculate voxel-level absorbed doses in nuclear medicine.We compared DOSXYZnrc Monte Carlo (MC) with published voxel-S-values to establish MC as truth. GBBS was investigated for mono-energetic 1.0, 0.1, and 0.01 MeV electron and photon sources as well as 131I and 90Y radionuclides. We investigated convergence of GBBS by analyzing different meshes (), energy group structures () for each radionuclide component, angular quadrature orders (, and scattering order expansions (–); higher indices imply finer discretization. We compared GBBS to MC in (1) voxel-S-value geometry for soft tissue, lung, and bone, and (2) a source at the interface between combinations of lung, soft tissue, and bone.Excluding Auger and conversion electrons, MC agreed within ≈5% of published source voxel absorbed doses. For the finest discretization, most GBBS absorbed doses in the source voxel changed by less than 1% compared to the next finest discretization along each phase space variable indicating sufficient convergence. For the finest discretization, agreement with MC in the source voxel ranged from −3% to −20% with larger differences at lower energies (−3% for 1 MeV electron in lung to −20% for 0.01 MeV photon in bone); similar agreement was found for the interface geometries. Differences between GBBS and MC in the source voxel for 90Y and 131I were −6%.The GBBS ATTILA was benchmarked against MC in the nuclear medicine regime. GBBS can be a viable alternative to MC for voxel-level absorbed doses in nuclear medicine. However, reconciliation of the differences between GBBS and MC at lower energies requires further investigation of energy deposition cross-sections.

Verification of I-125 brachytherapy source strength for use in radioactive seed localization procedures

A general-purpose nuclear medicine dose calibrator was assessed as a potential replacement for a dedicated air-communicating well-type ionization chamber (brachytherapy source strength verification instrument) for 125I seed source strength verification for radioactive seed localization, where less stringent accuracy tolerances may be acceptable. The accuracy, precision and reproducibility of the dose calibrator were measured and compared to regulatory requirements. The results of this work indicate that a dose calibrator can be used for 125I seed source strength verification for radioactive seed localization.

EYE LENS DOSES IN NUCLEAR MEDICINE: A MULTICENTRIC STUDY IN BELGIUM AND POLAND.

This study aimed to investigate the level of the eye lens (EL) doses in nuclear medicine in the light of the new International Commission on Radiological Protection limit. In 7 Belgian and 1 Polish hospitals, 45 staff members were monitored for EL (Hp(3)) and whole-body (WB) (Hp(10)) doses using dedicated dosemeters. Weekly measurements were carried out and used to estimate annual doses. Mostly diagnostic procedures involving radionuclides such as (99m)Tc and (18)F were monitored; measurements were also performed for therapeutic procedures. The cumulative doses showed important variation across the participants. The weekly EL and WB doses ranged from 0.02 to 0.27 and 0.03 to 0.17 mSv, respectively; the annual EL and WB doses ranged from 0.6 to 9.3 and 0.9 to 8.0 mSv, respectively. Some correlation was found between the EL and the WB doses. No significant correlation with the manipulated activities was found.

Use of PET/CT instead of CT-only when planning for radiation therapy does not notably increase life years lost in children being treated for cancer.

PET/CT may be more helpful than CT alone for radiation therapy planning, but the added risk due to higher doses of ionizing radiation is unknown. To estimate the risk of cancer induction and mortality attributable to the [F-18]2-fluoro-2-deoxyglucose (FDG) PET and CT scans used for radiation therapy planning in children with cancer, and compare to the risks attributable to the cancer treatment. Organ doses and effective doses were estimated for 40 children (2-18years old) who had been scanned using PET/CT as part of radiation therapy planning. The risk of inducing secondary cancer was estimated using the models in BEIR VII. The prognosis of an induced cancer was taken into account and the reduction in life expectancy, in terms of life years lost, was estimated for the diagnostics and compared to the life years lost attributable to the therapy. Multivariate linear regression was performed to find predictors for a high contribution to life years lost from the radiation therapy planning diagnostics. The mean contribution from PET to the effective dose from one PET/CT scan was 24% (range: 7-64%). The average proportion of life years lost attributable to the nuclear medicine dose component from one PET/CT scan was 15% (range: 3-41%). The ratio of life years lost from the radiation therapy planning PET/CT scans and that of the cancer treatment was on average 0.02 (range: 0.01-0.09). Female gender was associated with increased life years lost from the scans (P < 0.001). Using FDG-PET/CT instead of CT only when defining the target volumes for radiation therapy of children with cancer does not notably increase the number of life years lost attributable to diagnostic examinations.

Dose equivalent rate constants and barrier transmission data for nuclear medicine facility dose calculations and shielding design.

A primary goal of nuclear medicine facility design is to keep public and worker radiation doses As Low As Reasonably Achievable (ALARA). To estimate dose and shielding requirements, one needs to know both the dose equivalent rate constants for soft tissue and barrier transmission factors (TFs) for all radionuclides of interest. Dose equivalent rate constants are most commonly calculated using published air kerma or exposure rate constants, while transmission factors are most commonly calculated using published tenth-value layers (TVLs). Values can be calculated more accurately using the radionuclide's photon emission spectrum and the physical properties of lead, concrete, and/or tissue at these energies. These calculations may be non-trivial due to the polyenergetic nature of the radionuclides used in nuclear medicine. In this paper, the effects of dose equivalent rate constant and transmission factor on nuclear medicine dose and shielding calculations are investigated, and new values based on up-to-date nuclear data and thresholds specific to nuclear medicine are proposed. To facilitate practical use, transmission curves were fitted to the three-parameter Archer equation. Finally, the results of this work were applied to the design of a sample nuclear medicine facility and compared to doses calculated using common methods to investigate the effects of these values on dose estimates and shielding decisions. Dose equivalent rate constants generally agreed well with those derived from the literature with the exception of those from NCRP 124. Depending on the situation, Archer fit TFs could be significantly more accurate than TVL-based TFs. These results were reflected in the sample shielding problem, with unshielded dose estimates agreeing well, with the exception of those based on NCRP 124, and Archer fit TFs providing a more accurate alternative to TVL TFs and a simpler alternative to full spectral-based calculations. The data provided by this paper should assist in improving the accuracy and tractability of dose and shielding calculations for nuclear medicine facility design.