Radionuclides Of Interest Research Articles

Activation cross-sections of proton-induced nuclear reactions on natural zinc in the energy range of 4–30 MeV

In the present work, the excitation functions for the proton-induced nuclear reactions on natural zinc were measured in the energy range of 4–30 MeV using the well-established stacked-foil activation procedure. The activation products were measured based on their characteristic gamma lines using HPGe γ-ray spectrometry. The radioactivities determined were used for the calculation of cross-sections of the radionuclides of interest, i.e., 61Cu, 62,65,69mZn, 57Co and 66,67,68Ga. The cross-sections have also been compared with the available literature data and the theoretical prediction of the TALYS code via its latest TENDL-2023 library as well as the prediction of EMPIRE-3.2.2 model code. The results show a reasonable agreement when compared with the available literature data. This work, however, shows that the theoretical data extracted from the TENDL-2023 library and EMPIRE-3.2.2 default calculation significantly underestimate the experimental cross-sections of 57Co nuclide while they overestimate those of 69mZn. The present result has potential applications to improve the predicting capability of the model codes as well as to serve as additional data for the nuclear reaction cross-section database.

The production and separation of 161Tb with high specific activity at the University of Utah

Targeted radiotherapy (TRT) is an increasingly prominent area of research in nuclear medicine, particularly in the context of treating cancerous tumors. One radionuclide of considerable interest for TRT is terbium-161 (t1/2 = 6.95 days), which undergoes beta emission and shares similar decay properties as 177Lu (FDA-approved as LUTATHERA® and PLUVICTO®). Besides beta emission, 161Tb also emits a significant number of conversion and Auger electrons further enhancing its therapeutic potential. Terbium-161 can be produced using nuclear reactors through an indirect neutron capture reaction, G64160dn,γG64161d→3.66min,β−T65161b, from 160Gd targets. However, a key challenge in utilizing 161Tb for TRT lies in effectively separating target and product materials to attain high specific activity for radiolabeling. Here, we detail the production of no-carrier added 161Tb using low flux research reactors (mean thermal (<0.625 eV) neutron flux: 1.356×1012n∙cm−2∙s−1) like the University of Utah TRIGA Reactor, using enriched 160Gd2O3 targets (1.5 ± 0.3 μCi of 161Tb per mg of 160Gd target per hour of irradiation). We also developed a separation technique based on cation exchange and extraction chromatography, suitable for mCi level irradiations with targets exceeding 200 mg. In a simulated full-scale irradiation, 161Tb was successfully isolated from large mass targets using cation exchange (AG 50W-X8, with 2-hydroxyisobutyric acid at 70 mM, pH 4.75) and extraction chromatography (LN Resin, 0.5–0.75 M HNO3) methods. This resulted in high apparent molar activities of [161Tb]Tb-DOTA (113 ± 3 MBq/nmol), demonstrating high purity 161Tb relevant for potential future preclinical applications.

Assessment of environmental impacts from authorized discharges of tritiated water from the Fukushima site to coastal and offshore regions

In August 2023, the long-planned discharging of radioactive wastewater from the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) started after the confirmation of its feasibility and safety. As this water contains elevated amounts of tritium even after being diluted, a lot of resources have been invested in the monitoring of the Fukushima coastal region where the discharge outlet is located. We compare the first 3H surface activity concentrations from these measurements (up to the end of November 2023) with the available background values to evaluate a possible impact of the long-term discharging on humans and environmental levels of the radionuclide of interest in the same or nearby area. From our results, we can conclude that the joint effect of horizontal and vertical mixing has been significant enough to reduce tritium concentrations at the monitored locations in the region close to the FDNPP port two days after the end of the respective phase of the discharging beyond the detection limit of the applied analytical methods (∼0.3 Bq L−1) which is by five orders of magnitude lower than safety limit for drinking water set by the World Health Organization (WHO). Moreover, the distant correlation analysis showed that tritium concentrations at stations located further than 1.4 km were very close to pre-discharge levels (∼0.4 Bq L−1). We also estimated that the 3H activity concentration in the offshore Fukushima region would be elevated by 0.01 Bq L−1 at maximum over a year of continuous discharging, which is in concordance with the already published modeling papers and much less than the impact of the FDNPP accident in 2011.

An electrochemical generator for the continual supply of 213Bi from 225Ac for use in targeted alpha therapy applications

Bismuth-213 is a radionuclide of interest for targeted alpha therapy and is supplied via a radiochemical generator system through the decay of 225Ac. Radionuclide generators employ longer lived "parent" radionuclides to routinely supply shorter-lived "daughter" radionuclides. The traditional 225Ac/213Bi radiochemical generator relies on an organic cation exchange resin where 225Ac binds to the resin and 213Bi is routinely eluted. These resins degrade when they absorb large doses of ionizing radiation (>1×106Gy/mg), which has been observed when the loading activity of 225Ac exceeds 2.59*109Bq (70mCi). Herein we report the development of an electrochemical generator for the supply of 213Bi that has the potential to overcome this limitation. Bismuth-213 spontaneously electrodeposits onto nickel foils in 0.1M hydrochloric acid at 70°C. Using this method, we were able to plate an average of 73±4% of the 213Bi in solution and obtain a final 213Bi recovery of 65±8% in 0.1M citrate pH4.5 via reverse electrolysis using titanium as the cathode. The recovered 213Bi had an average radiochemical purity of >99.8% and was successfully used to radiolabel DOTATATE with an average radiochemical yield of 85.1% (not optimized).

Naturally occurring radioactive material (NORM) in Saudi Arabia: A Review

Saudi Arabia, the world's largest crude oil exporter, also possesses rich deposits of natural resources like phosphate, bauxite, copper, and various minerals. Extracting and processing these resources can concentrate naturally occurring radionuclides in the resulting products, byproducts, residues, and wastes. This can sometimes elevate their activity concentrations above original levels, potentially increasing exposure risks for workers and the public, and contaminating the environment.This review provided a comprehensive overview of naturally occurring radioactive materials (NORM) across various industries in Saudi Arabia. It outlines the regulatory framework established by the country's authority for NORM management. By reviewing current literature, the article presents activity concentrations of key radionuclides of interest - 40K and those from the 238U and 232Th decay series-within different industries. The review highlights the need for further research to assess the impact of industrial processes on the activity concentrations of natural radionuclides. It concludes by calling for additional studies to address the significant knowledge gap on NORM in specific industries, particularly phosphate, bauxite, water treatment plant and oil and gas, which have the potential to generate radioactive waste. The importance of characterizing industries involving NORM is emphasized for evaluating radiological impact and implementing appropriate safety measures.

Measurements of the absolute gamma-ray emission intensities from the decay of 166Ho

166Ho (T1/2≈ 26.8 h) is an emerging theragnostic radionuclide of interest in nuclear medicine due to its peculiar decay scheme, featuring high-energy β− emission (≈ 1.8 MeV) coupled with the main gamma-ray emission (≈ 80.6 keV). Using the new 166Ho activity standard and the well-calibrated, high-energy resolution HPGe detector, both available at ENEA-INMRI, a new determination of several 166Ho gamma-ray emission intensities, Iγ, was performed with low uncertainty. The new Iγ values contributed to the Decay Data Evaluation Project.

Selection of radionuclide(s) for targeted alpha therapy based on their nuclear decay properties.

Targeted alpha therapy (TAT) is a promising form of oncology treatment utilising alpha-emitting radionuclides that can specifically accumulate at disease sites. The high energy and high linear energy transfer associated with alpha emissions causes localised damage at target sites whilst minimising that to surrounding healthy tissue. The lack of appropriate radionuclides has inhibited research in TAT. The identification of appropriate radionuclides should be primarily a function of the radionuclide's nuclear decay properties, and not their biochemistry or economic factors since these last two factors can change; however, the nuclear decay properties are fixed to that nuclide. This study has defined and applied a criterion based on nuclear decay properties useful for TAT. This down-selection exercise concluded that the most appropriate radionuclides are: 149 Tb, 211 At/ 211 Po, 212 Pb/ 212 Bi/ 212 Po, 213 Bi/ 213 Po, 224 Ra, 225 Ra/ 225 Ac/ 221 Fr, 226 Ac/ 226 Th, 227 Th/ 223 Ra/ 219 Rn, 229 U, 230 U/ 226 Th, and 253 Fm, the majority of which have previously been considered for TAT. 229 U and 253 Fm have been newly identified and could become new radionuclides of interest for TAT, depending on their decay chain progeny.

Self-characterization of irradiation facility using synthetic multi-element standard for determination of k0-factors of seven radionuclides of interest

In this study, the k0 factors of 7 short and medium-lived radionuclides (i.e. 66Cu, 52V, 38Cl, 134mCs, 128I, 140La and 56Mn) were experimentally determined at the Dalat nuclear research reactor (DNRR). The synthetic multi-element standard (SMELS) samples were irradiated at Channel 7-1 installed at the DNRR. SMELS Type III contains Au and Zr, thus also allowing the self-characterization of the irradiation facility (i.e. the direct determination of f and a with only this material). The k0 factors of 7 short and medium-lived radionuclides obtained in this work are a good agreement with the reference k0 factors. The ratio values between the experimental and reference k0 values were 66Cu of 1.004, 52V of 1.004, 38Cl of 0.961 and 0.976 (1642.7 keV and 2167.4keV, respectively), 134mCs of 0.831, 128I of 1.022, 140La of 0.936, 0.955 and 0.955 (328.8 keV, 487.0 keV and 1596.2keV, respectively), 56Mn of 0.985 and 0.973 (846.8 keV and 1810.7keV, respectively). The radionuclides range of uncertainty values were lower than 7%. The present result indicates that the k0 factors can be determined by using synthetic multi-element standards (SMELS).

Terbium radionuclides for theranostic applications in nuclear medicine: from atom to bedside.

Terbium features four clinically interesting radionuclides for application in nuclear medicine: terbium-149, terbium-152, terbium-155, and terbium-161. Their identical chemical properties enable the synthesis of radiopharmaceuticals with the same pharmacokinetic character, while their distinctive decay characteristics make them valuable for both imaging and therapeutic applications. In particular, terbium-152 and terbium-155 are useful candidates for positron emission tomography (PET) and single photon emission computed tomography (SPECT) imaging, respectively; whereas terbium-149 and terbium-161 find application in α- and β--/Auger electron therapy, respectively. This unique characteristic makes the terbium family ideal for the "matched-pair" principle of theranostics. In this review, the advantages and challenges of terbium-based radiopharmaceuticals are discussed, covering the entire chain from radionuclide production to bedside administration. It elaborates on the fundamental properties of terbium, the production routes of the four interesting radionuclides and gives an overview of the available bifunctional chelators. Finally, we discuss the preclinical and clinical studies as well as the prospects of this promising development in nuclear medicine.

Ultra-low background germanium assay at the Boulby Underground Laboratory

As we move to an era where next generation ultra-low background particle physics experiments begin to be designed and constructed, the ability to assay materials with high sensitivity and at speed with a variety of techniques will be key. This paper describes the Mirion Technologies (Canberra) specialty ultra-low background detectors installed and commissioned at the Boulby Underground Laboratory between 2017 and 2021. The low background levels of the detectors combine with low background shielding and a radon-reduced dry nitrogen purge system to give sensitivity approaching the best in the world without the need for intricate shielding solutions. For an optimised sample geometry, run for 100 d, it would be possible to reach close to 10 μBq kg-1 (10-12 g/g) for background radionuclides of interest in neutrinoless double-beta decay.

Nuclear Cross-Section of Proton-Induced Reactions on Enriched 48Ti Targets for the Production of Theranostic 47Sc Radionuclide, 46cSc, 44mSc, 44gSc, 43Sc, and 48V.

The cross-sections of the 48Ti(p,x)47Sc, 46cSc, 44mSc, 44gSc, 43Sc, and 48V nuclear reactions were measured from 18 to 70 MeV, with particular attention to 47Sc production. Enriched 48Ti powder was deposited on an aluminum backing and the obtained targets were characterized via elastic backscattering spectroscopy at the INFN-LNL. Targets were exposed to low-intensity proton irradiation using the stacked-foils technique at the ARRONAX facility. Activated samples were measured using γ-spectrometry; the results were compared with the data int he literature and the theoretical TALYS-based values. A regular trend in the new values obtained from the different irradiation runs was noted, as well as a good agreement with the literature data, for all the radionuclides of interest: 47Sc, 46cSc, 44mSc, 44gSc, 43Sc, and 48V. 47Sc production was also discussed, considering yield and radionuclidic purity, for different 47Sc production scenarios.

A combined approach for the calculation of activation yields and the characterization of materials for a medical cyclotron

Cyclotrons for the production of radiopharmaceuticals have become important tools in modern nuclear medicine. At the end of their lifecycles, such installations have to be dismantled and any activated materials must be treated according to the local radiation protection legislation. Using a simulation model, we have developed a non-destructive approach for the radiological characterization of components inside and around an IBA Cyclone 18/9 cyclotron. The methodology is based on software tools developed at CERN (FLUKA Monte Carlo code, ActiWiz and RAW). The simulation results were compared to measurements made using reference samples placed around the cyclotron inside the bunker. Results show a reasonable agreement between simulation and measurements of about a factor of two for a set of 27 reference samples and 11 radionuclides of interest. The origin of this factor has been thoroughly evaluated and opened the door to further investigations leading to possible avenues for improvement.

Comparison of methods to determine extraction efficiencies of Ra isotopes and 227Ac from large volume seawater samples

Radium isotopes, other than 226Ra, and 227Ac are typically present at low activities in the open ocean. The analysis of these isotopes thus requires the collection of large volumes of seawater and high sensitivity, low background instruments. To obtain the required large volumes (hundreds to thousands of liters), these radionuclides are typically preconcentrated on cartridge-style filters impregnated with MnO2 (Mn-cartridges) deployed on in-situ pumps. This technique, however, requires the determination of the extraction efficiency of the Mn-cartridges for the radionuclides of interest. For Ra isotopes, we used two methods to estimate the extraction efficiency of these Mn-cartridges at two stations on the South-West Indian Ridge in the Southern Ocean (GEOTRACES GS02). Method (1) compares the 226Ra activities recovered on the Mn-cartridges versus the activities determined in Mn-fibers, through which seawater was passed at a flow rate < 1 L min−1 to quantitatively sorb Ra (Mn-fiber method) while method (2) combines the 226Ra activities determined from two Mn-cartridges placed in series on in-situ pumps (A-B method). The second method is also applied to determine the 227Ac extraction efficiency. We find a relatively wide-range of Ra and 227Ac extraction efficiencies across the dataset (from 44.8% to 99.6% for Ra, and from 23.7% to 77.5% for 227Ac). Overall, the yield of 227Ac extraction is lower than that of Ra (mean value of 49.3 ± 19.0% for 227Ac, n = 10, mean value of 79.2 ± 10.3% for Ra, n = 13, using the Mn-fiber method; and a mean value of 63.9 ± 12.5%, n = 11 using the A-B method). Our dataset suggests that the Ra extraction efficiencies using either the A-B method or the Mn-fiber method are in relatively good agreement. Consequently, the 223Raex, 224Raex and 228Ra activities determined from the Mn-cartridges by applying the two Ra extraction yields are similar. We also show that the 227Ac extraction efficiency can be estimated from the Ra extraction efficiency allowing the use of a single Mn-cartridge. Finally, we recommend to determine the Ra and 227Ac extraction efficiencies in each individual Mn-cartridge, rather than applying a single extraction efficiency to all the Mn-cartridges, since a significant variability in the extraction efficiencies was observed between the different Mn-cartridges.

Allometric-kinetic model predictions of concentration ratios for anthropogenic radionuclides across animal classes and food selection

Protection of the environment from radiation fundamentally relies on dose assessments for non-human biota. Many of these dose assessments use measured or predicted concentrations of radionuclides in soil or water combined with Concentration Ratios (CRs) to estimate whole body concentrations in animals and plants, yet there is a paucity of CR data relative to the vast number of potential taxa and radioactive contaminants in the environment and their taxon-specific ecosystems. Because there are many taxa each having very different behaviors and biology, and there are many possible bioavailable radionuclides, CRs have the potential to vary by orders-of-magnitude, as often seen in published data. Given the diversity of taxa, the International Commission on Radiological Protection (ICRP) has selected 12 non-human biota as reference animals and plants (RAPs), while the U.S. Department of Energy (DOE) uses the non-taxon specific categories of terrestrial, riparian, and aquatic animals. The question we examine here, in part, is: are these RAPs and categorizations sufficient to adequately protect all species given the broad diversity of animals in a region? To explore this question, we utilize an Allometric-Kinetic (A-K) model to calculate radionuclide-specific CRs for common animal classes, which are then further subcategorized into herbivores, omnivores, carnivores, and invertebrate detritivores. Comparisons in CRs among animal classes exhibited only small differences, but there was order of magnitude differences between herbivores, carnivores, and especially detritivores, for many radionuclides of interest. These findings suggest that the ICRP RAPs and the DOE categories are reasonable, but their accuracy could be improved by including sub-categories related to animal dietary ecology and biology. Finally, comparisons of A-K model predicted CR values to published CRs show order-of-magnitude variations, providing justification for additional studies of animal assimilation across radionuclides, environmental conditions, and animal classes.

Experimental assessment of nuclear cross sections for the production of Tb radioisotopes with a medical cyclotron

155Tb is one of the most interesting radionuclides for theranostic applications. It is suitable for SPECT imaging and it can be used as a true diagnostic partner of the therapeutic 149Tb and 161Tb. Its production by proton irradiation using enriched 155Gd and 156Gd oxide targets is currently being investigated and represents a promising solution. To achieve the level of radionuclidic purity required in the clinical setting, the co-production of Tb impurities has to be minimized. For this purpose, an accurate knowledge of the cross sections of the nuclear reactions involved is of paramount importance. In this paper, we report on the assessment of cross sections of the reactions 154Gd(p,xn)153,154,154m1,154m2Tb, 155Gd(p,xn)154,154m1,154m2,155Tb, 156Gd(p,xn)155,156Tb and 157Gd(p,2n)156Tb derived with a specific data analysis procedure developed by our group. This method allows to disentangle the nuclear contributions from the production cross section by inverting linear systems of equations and it requires the measurement of the cross sections from as many materials as the reactions involved in the production of the radionuclide under study. For this purpose, the experimental data previously measured by our group at the Bern medical cyclotron by irradiating natural Gd2O3, enriched 155Gd2O3 and enriched 156Gd2O3 targets were used. For some of these nuclear reactions, cross sections were assessed for the first time. On the basis of our findings, production yield and purity can be calculated for any kind of isotopic composition of the enriched material.

161Tb-DOTATOC Production Using a Fully Automated Disposable Cassette System: A First Step Toward the Introduction of 161Tb into the Clinic.

161Tb is an interesting radionuclide for application in the treatment of neuroendocrine neoplasms' small metastases and single cancer cells because of its conversion and Auger-electron emission. Tb has coordination chemistry similar to that of Lu; therefore, like 177Lu, it can stably radiolabel DOTATOC, one of the leading peptides used for the treatment of neuroendocrine neoplasms. However, 161Tb is a recently developed radionuclide that has not yet been specified for clinical use. Therefore, the aim of the current work was to characterize and specify 161Tb and to develop a protocol for the synthesis and quality control of 161Tb-DOTATOC with a fully automated process conforming to good-manufacturing-practice guidelines, in view of its clinical use. Methods: 161Tb, produced by neutron irradiation of 160Gd in high-flux reactors followed by radiochemical separation from its target material, was characterized regarding its radionuclidic purity, chemical purity, endotoxin level, and radiochemical purity (RCP) in analogy to what is described in the European Pharmacopoeia for no-carrier-added 177Lu. In addition, 161Tb was introduced into a fully automated cassette-module synthesis to produce 161Tb-DOTATOC, as used for 177Lu-DOTATOC. The quality and stability of the produced radiopharmaceutical in terms of identity, RCP, and ethanol and endotoxin content were assessed by means of high-performance liquid chromatography, gas chromatography, and an endotoxin test, respectively. Results: 161Tb produced under the described conditions showed, as the no-carrier-added 177Lu, a pH of 1-2, radionuclidic purity and RCP of more than 99.9%, and an endotoxin level below the permitted range (175 IU/mL), indicating its appropriate quality for clinical use. In addition, an efficient and robust procedure for the automated production and quality control of 161Tb-DOTATOC with clinically applicable specifications and activity levels, that is, 1.0-7.4 GBq in 20 mL, was developed. The radiopharmaceutical's quality control was also developed using chromatographic methods, which confirmed the product's stability (RCP ≥ 95%) over 24 h. Conclusion: The current study demonstrated that 161Tb has appropriate features for clinical use. The developed synthesis protocol guarantees high yields and safe preparation of injectable 161Tb-DOTATOC. The investigated approach could be translated to other DOTA-derivatized peptides; thus, 161Tb could be successfully applied in clinical practice for radionuclide therapy.

Application of multiple source preparation procedures on environmental samples for determination of Uranium in Alpha Spectrometry

Different source preparation procedures for determination of uranium radioisotopes in environmental samples by alpha spectrometry are presented. Addition of 232U tracer was followed by the preconcentration of radionuclides of interest from the samples by co-precipitation on Fe(OH)3 at pH 9 using ammonia solution. The uranium extracts were obtained by column separation using UTEVA resin after which the uranium sources from the samples were prepared using either electrodeposition or microcoprecipitation with LaF3 before counting. The procedures were tested on spinach, soil, tap water and waste water from the Libanon Gold mine and the results obtained were compared.

Decay spectroscopy of the blocked fission product I130

Numerous applications rely on the identification and quantification of fission products with the activation technique, where $\ensuremath{\gamma}$ rays emitted in the decay are used to estimate the initial activity of the radionuclide of interest. $^{130}\mathrm{I}$ is a so-called blocked fission product, which can be produced only directly through fission, a property that makes it particularly attractive for nuclear forensics. A source of $^{130}\mathrm{I}$ was produced using a $(p,n)$ reaction on enriched $^{130}\mathrm{Te}$ at the Brookhaven Tandem Van de Graaff, and its decay was studied with the Gammasphere at Argonne National Laboratory. Two new levels were identified, and over 25 transitions were added, removed, or replaced in the level scheme with intensity measurements made down to ${I}_{\ensuremath{\gamma}}$ = 0.000 66 per 100 decays. The uncertainty on the intensities of the strongest transitions, those that are commonly used to quantify the activity of the radionuclide, was improved by a factor of 2 compared to the previous best assessment, and discrepancies in the literature values were resolved. A detailed angular correlation analysis further permitted the determination of a number of spin assignments for excited levels and mixing ratios for $\ensuremath{\gamma}$-ray transitions.

Eye-lens dose rate conversion factors due to hot particles and surface contaminations on the cornea

In 2012, the International Commission on Radiological Protection issued new recommendations, in publication 118, regarding the dose limits to the eye-lens. New analyses of historical exposure data had indicated that radiation-induced cataracts may appear at lower doses than previously assumed. This spurred largescale efforts in a variety of fields including dosimetry, radiation effects simulations, and the review of national regulatory limits. On the simulation side, much work led to the publication of dose rate conversion factors (DRCFs), to calculate the dose to the radiosensitive part of the eye-lens, and to the whole eye-lens as functions of the incident fluence of electron, photon, positron, and neutron radiation. The standard, ISO-15382 (2015 Radiological Protection—Procedures for Monitoring the Dose to the Lens of the Eye, the Skin and the Extremities), from the International Organization for Standardization (ISO), stated that the direct contact of a hot radioactive particle on the eye-lens represents a special contamination condition that must be considered. The aim of this work was to produce tabulated data of eye-lens dose rates, per activity (MBq), for a variety of radionuclides. In this work, the dose to the eye-lens from contamination directly in contact with the cornea, expressed in terms of DRCFs for eye-lens, in units of Gy h−1 MBq−1, are presented for 102 radionuclides of interest. These radionuclides were selected as they had been considered by the International Atomic Energy Agency of importance for skin dose. The method consisted of two steps. The first was the determination of the DRCFs for mono-energetic electrons and photons for a hot particle in contact with the eye-lens, followed by the folding of these quantities with the emissions of the radionuclides of interest. Contributions from spontaneous fission neutrons were considered separately. Exposure geometries for spherical hot particles of different dimensions, materials and locations on the cornea were considered. In addition, partial surface coverage of the cornea, consistent with an accidental exposure to a contaminated liquid, was also modelled. Resulting radionuclide DRCFs were verified, for a few specific geometries and radionuclides with dedicated Monte Carlo simulations. The final data are presented in several tables included in this paper.

ESTIMATION OF PROTECTION FACTORS FOR SURROGATE VEHICLE MODELS IN ENVIRONMENTAL CONTAMINATION SCENARIOS.

In the event of a nuclear or radiological incident involving the release of radioactive material, it may be critical to estimate doses to individuals who are transported through contaminated areas by vehicles. The radiation protection factor (RPF) for vehicles can be calculated to determine the level of shielding protection by the vehicle from external radiation sources. Prior studies evaluating RPFs demonstrate a lack of realistic vehicle configurations and the results cannot be extended directly to scenarios where a vehicle is surrounded by a contaminated environmental field. In this work, sex-averaged effective dose-rate coefficients were computed employing International Commission on Radiological Protection Publication 103 recommendations for radionuclides of interest and used to calculate the RPF as the ratio of unshielded to shielded dose for various radionuclides of interest in consequence management scenarios. Comparisons to dose reduction factors calculated using environmental measurement data from the 2011 Fukushima Daiichi nuclear incident were conducted to benchmark to experimental field measurements.