Ambient Dose Research Articles

Analysis of the relationship between ambient dose, ambient dose equivalent and effective dose in operational neutron spectra.

This study examines the relationship between ambient dose H*, ambient dose equivalent H*(10), and effective dose E for external neutron irradiation over 163 operational spectra from different workplaces. The results show that H* provides a reasonable estimate with a controlled margin, even if overestimated, to assess E compared with H*(10), which can lead to a significant overestimation or underestimation of E depending on the neutron spectra. The results highlight the limitations of H*(10) and the superiority of H* in estimating E according to the requirements of the operational quantities, especially in environments with high-energy neutrons.

Radiological characterization of the tailings of an abandoned copper mine using a neural network and geostatistical analysis through the Co-Kriging method

The radiological characterization of soil contaminated with natural radionuclides enables the classification of the area under investigation, the optimization of laboratory measurements, and informed decision-making on potential site remediation. Neural networks (NN) are emerging as a new candidate for performing these tasks as an alternative to conventional geostatistical tools such as Co-Kriging. This study demonstrates the implementation of a NN for estimating radiological values such as ambient dose equivalent (H*(10)), surface activity and activity concentrations of natural radionuclides present in a waste dump of a Cu mine with a high level of natural radionuclides. The results obtained using a NN were compared with those estimated by Co-Kriging. Both models reproduced field measurements equivalently as a function of spatial coordinates. Similarly, the deviations from the reference concentration values obtained in the output layer of the NN were smaller than the deviations obtained from the multiple regression analysis (MRA), as indicated by the results of the root mean square error. Finally, the method validation showed that the estimation of radiological parameters based on their spatial coordinates faithfully reproduced the affected area. The estimation of the activity concentrations was less accurate for both the NN and MRA; however, both methods gave statistically comparable results for activity concentrations obtained by gamma spectrometry (Student's t-test and Fisher's F-test).

New guidelines on radiation survey and sanitary assessment of residential, public and industrial buildings and facilities in terms of radiation safety indicators. Part 1

The paper (in two parts) presents an overview of the new guidelines MR 2.6.1.0333-23 (approved on 01 December 2023) that supersede guidelines MU 2.6.1.2838-11, which were used for organizing radiation surveys of residential, public and industrial buildings and facilities and their sanitary assessment in terms of radiation safety indicators over the past 12 years. Due to a large number of critical comments on the document received during this period, a need for significant revision emerged. The scope of the revised document was expanded and now covers all stages of the life cycle of buildings and facilities: commissioning; operation period; overhaul and reconstruction; demolition. In the first part of the paper, the structure and legal status of the new guidelines are considered, as well as the main innovations in terms of measuring the ambient dose equivalent rate of gamma radiation, searching and identifying local radiation anomalies and sites of radioactive contamination, measuring surface contamination levels, and determining the minimum number of premises to be surveyed. The reasons for the introduction of certain changes are outlined. The issues of evaluation of measurement uncertainty, reporting of measurement results and registration of test reports are covered in detail. In the second part of the paper, numerous innovations will be considered in terms of estimating the indoor average annual equilibrium equivalent concentration of radon isotopes.

Radiation survey of the territory of Teriberka village and Kildin Island of Murmansk Region

The Arctic zone is a strategically important territory of the Russian Federation from the point of view of socio-economic, scientific, and technological components, as well as in terms of planning the development of a system for ensuring national defense, state, and public security. Intensive exploration and development of Arctic territories, combined with a rapidly changing environment due to global climate change, carries a few risks for the environment. This determines the need to assess potential environmental damage and determine the significance of factors of potential negative impact on unique Arctic ecosystems now and in the future. The article presents current data from a radiation survey of the coastal zone of residential areas close to the site of the sinking of the nuclear submarine ‘K-159’ – Teriberka village and Kildin Island, Murmansk region, as well as the sea area to assess the primary levels of man-made and natural radionuclides before starting work on the potential lifting of the submerged object. This work was carried out as part of the implementation of the State Program of the Russian Federation ‘Health Development’ (approved by Decree of the Government of the Russian Federation dated December 26, 2017, No. 1640). The results of measuring the ambient dose equivalent rate of gamma radiation in open areas and determining the content of man-made and natural radionuclides in samples of environmental media (soil, aquatic vegetation, marine aquatic organisms) and food products of local origin and production (mushrooms, berries, fish, milk). The research results showed that the state of the study area is generally stable in terms of the radiation factor, the content of radionuclides is at the background level. As part of this work, electronic database modules have been developed based on the results of radiation monitoring in Teriberka village and on Kildin Island. Certificate No. 2022621890 was received on registration of a database with the results of the radiation situation of the above-mentioned territories.

A high-throughput focused collimator for OAR-sparing preclinical proton FLASH studies: commissioning and validation

Objective. To fabricate and validate a novel focused collimator designed to spare normal tissue in a murine hemithoracic irradiation model using 250 MeV protons delivered at ultra-high dose rates (UHDRs) for preclinical FLASH radiation therapy (FLASH-RT) studies. Approach. A brass collimator was developed to shape 250 MeV UHDR protons from our Varian ProBeam. Six 13 mm apertures, of equivalent size to kV x-ray fields historically used to perform hemithorax irradiations, were precisely machined to match beam divergence, allowing concurrent hemithoracic irradiation of six mice while sparing the contralateral lung and abdominal organs. The collimated field profiles were characterized by film dosimetry, and a radiation survey of neutron activation was performed to ensure the safety of staff positioning animals. Main results. The brass collimator produced 1.2 mm penumbrae radiation fields comparable to kV x-rays used in preclinical studies. The penumbrae in the six apertures are similar, with full-width half-maxima of 13.3 mm and 13.5 mm for the central and peripheral apertures, respectively. The collimator delivered a similar dose at an average rate of 52 Gy s−1 for all apertures. While neutron activation produces a high (0.2 mSv h−1) initial ambient equivalent dose rate, a parallel work-flow in which imaging and setup are performed without the collimator ensures safety to staff. Significance. Scanned protons have the greatest potential for future translation of FLASH-RT in clinical treatments due to their ability to treat deep-seated tumors with high conformality. However, the Gaussian distribution of dose in proton spots produces wider lateral penumbrae compared to other modalities. This presents a challenge in small animal pre-clinical studies, where millimeter-scale penumbrae are required to precisely target the intended volume. Offering high-throughput irradiation of mice with sharp penumbrae, our novel collimator-based platform serves as an important benchmark for enabling large-scale, cost-effective radiobiological studies of the FLASH effect in murine models.

Radioactivity estimation of radioactive hotspots using a Compton camera and derivation of dose rates in the surrounding environment

At the Fukushima Daiichi Nuclear Power Station, radiation sources released in the accident were deposited on various equipment and building structures. During decommissioning, it is crucial to understand the distribution of radiation sources and ambient dose equivalent rates to reduce worker exposure and implement detailed work planning. In this study, the author introduces a method for visualizing radiation sources, estimates their radioactivity using a Compton camera, and derives the dose rate around the radiation sources. In the demonstration test, the Compton camera was used to visualize radioactive hotspots caused by 137Cs radiation sources deposited in the outdoor environment and estimated the radioactivity. Furthermore, the dose rate around the hotspots was calculated from the estimated radioactivity, which confirmed that the calculated dose rate correlated with the dose rate measured using a survey meter. This approach is novel, where a series of analyses were conducted using the Compton camera to visualize radioactive hotspots, estimate the radioactivity, and derive the dose rate in the surrounding environment.

Photon and Neutron Dose Estimation Using Monte Carlo Simulation in TrueBeam’s Room

Purpose: The distribution of neutron ambient dose equivalent within the TrueBeam 10 MV photon chamber was investigated. Materials and Methods: The research used particle and heavy ion transport code system (PHITS) code and JENDL-5.0 to simulate the neutron ambient dose equivalent on and around TrueBeam’s head. The simulated results were compared with the measured results using CR-39 detectors when TrueBeam radiated 5000 monitor units of 10 MV photons with field sizes 20 cm × 20 cm and 0.5 cm × 0.5 cm. Results: Out of field size, the neutron ambient dose equivalents of the 0.5 cm × 0.5 cm field size are higher than those values of the 20 cm × 20 cm field size from 4% to 30%. The differences between the simulated value and the measured value of the neutron ambient dose equivalents at all points out of field size are smaller than 20%. Conclusion: The neutron ambient dose equivalents, simulated with PHITS and JENDL-5.0, are satisfied with the measured neutron ambient dose equivalent.

Exploration of deep-learning-based dose rate estimation model for a large-volume plastic scintillation detector

This study explores a deep-learning-based dose rate estimation model for a large-volume plastic scintillation detector. Large-volume plastic scintillators are typically utilized as detector materials for radiation portal monitors (RPMs); however, they can also be utilized for environmental radiation monitoring. We introduced a deep-learning approach to estimate the ambient dose equivalent (H*(10)) from the spectral input using a transformer architecture. The performance of the model was compared with that of other models with conventional architectures and verified using simulated and measured spectra. In addition, we present a model explanation and an uncertainty quantification approach to ensure the reliability of the dose rate estimation model. Our verification demonstrates that the behavior of the model aligns with domain knowledge and that the uncertainty of the model's estimations increases according to the increment in the statistical uncertainties of the input spectra. The experimental results indicate that the model shows a promising performance as a practical application for RPMs.

New concept for neutron dosimetry in ion beam radiotherapy: Feasibility study

In ion beam radiotherapy (IBRT), the secondary neutrons contribute to whole-body exposure, potentially increasing the risk of radiation-induced cancer. The energy range of neutrons generated in the IBRT treatment room spreads from thermal to a few hundred MeV, with those between 0.1 and a few hundred MeV playing a crucial role in the ambient dose equivalent. This study introduces a novel approach for measuring the ambient dose equivalent that focuses on the neutron energy spectrum in the IBRT treatment room. Distinguished by the use of a hydrogen-filled proportional counter and a single moderator, this new method overcomes the weakness of the traditional Bonner Sphere Spectrometer by reducing the complexity and time required for measurements. By deriving the neutron energy spectrum, the neutron ambient dose equivalent can be accurately calculated. Our validation demonstrated that the neutron ambient dose equivalents derived using our method closely agree with the actual values, with discrepancies within 4%, underscoring the feasibility of this innovative approach to neutron dosimetry in the IBRT treatment room.

Photon and neutron dose evaluation at the Beam Test Facility of the INFN - National Laboratory of Frascati

Dose evaluation and direct measurements are fundamental for radiation protection in non-conventional accelerator facilities, both before and after the primary and secondary shielding. In this paper, we will report about the experimental setup, data acquisition and analysis, together with FLUKA modeling, of the dose measurements test carried out in the Beam Test Facility (BTF) of the INFN - Frascati’s National Laboratories (LNF), where an intense mixed field is produced and measured with thermoluminescent dosimeters. BTF is an extraction and transport line of DAΦNE LINAC (Buonomo et al. 2021; Mazzitelli et al. 2003). It is optimized for electrons and positrons production in a wide range of intensity, energy (30 MeV–800 MeV), beam spot dimensions and divergence, using both primary and secondary beam of the DAΦNE LINAC. Through the years, the BTF has gained an important role in particle detectors test and development with electron/positron beam. A small fraction of the BTF’s shifts have been dedicated to radiation damage test using LINAC electron primary beam up to 5×1010 e-/s. As radiation protection group of the LNF, we evaluated the dose when electrons impinging on a Pb target from: (i) photon Bremsstrahlung production; (ii) photoneutron production. Three dedicated tests with 503 MeV electrons impinging on a ∼16 cm thick Pb target have been carried out in February, June 2022 and in January 2023, using TLD700 and TLD600, measuring doses at several charge intervals. The aim of this study focuses on evaluating dosimetric quantities produced by the mixed field, air kerma for the photon component, and ambient dose equivalent for the neutron one, using thermoluminescence dosimeters calibrated with low-energy standards: Cs-137 and Am-Be. The approach adopted involves the use of Monte Carlo simulations of the experiment, both to benchmark against experimental measurements and to validate the results obtained for energies higher than those of calibration. The results of this comparison show excellent agreement between measured and simulated quantities in the forward direction, allowing us to conclude and confirm the validity of the calibrations themselves.

Dose equivalent consideration from neutron contamination in modified radiotherapy vault: a Monte Carlo study

Laminated barriers incorporating metal sheets provide effective protection for space-restricted radiotherapy centers. This study aimed to assess photoneutron contamination in smaller vaults protected by different compositions of multilayer barriers during simulated pelvic radiotherapy with 18 MV photon beams. Monte Carlo Simulations of 18 MV LINAC (Varian 2100 C/D) and Medical Internal Radiation Dose (MIRD) phantom were used to assess photoneutron contamination within reconstructed vaults incorporating different combinations of metal sheet and borated polyethylene (BPE) during pelvic radiotherapy. The findings highlight a 3.27 and 2.91 times increase in ambient neutron dose Hn* (10) along the maze of reconstructed vaults that use lead and steel sheets, respectively, compared to concrete. The Hn* (10) outside the treatment room increased after incorporating a metal sheet, but it remained within the permissible limit of 20 μSv/week for uncontrolled areas adjacent to the LINAC bunker, even with a workload of 1000Gy/week. Neutron equivalent doses in the patient’s organs ranged from 0.22 to 0.96 mSv Gy−1. There is no notable distinction in the organ’s neutron equivalent dose, fatal cancer risk, secondary radiation-induced cancer risk, and cancer mortality for various laminated barrier compositions. Furthermore, the use of metal sheets for vault wall reconstruction keeps the variation in cancer risk induced by photoneutrons below 6%, while risks of fatal cancer and cancer mortality vary less than 11%. While the metal portion of the laminated barrier raises the neutron dose, the addition of a BPE plate reduces concerns of increased effective dose and secondary malignancy risk.

Ambient Radiological Condition around an Operating Uranium Mill Tailings Disposal Facility at Turamdih, India

AbstractLarge quantity of the feed forms part of the solid waste (tailings) slurry in the low grade uranium ore processing plant at Turamdih in Singhbhum region of Jharkhand, India. Coarser size fraction of this waste is used for backfilling of the underground uranium mines and fine fraction is discharged into an engineered impoundment system or tailings pond. Since inception of the discharges, field expeditions have been performed as part of the comprehensive radiological monitoring program for the facility. The monitoring program broadly consists of ambient gamma dose rate, atmospheric 222Rn and long lived alpha activity measurement around the facility and the estimation of radionuclide in diverse environmental matrices collected from the adjoining sites. Findings reflect that the radiological conditions are comparable to the pre-existing background level adjoining the tailings pond impoundment. Further, measurements for more than a decade confirm that the atmospheric radon concentration profile just a few meters beyond the embankment is indistinguishable from the background.

Оценка функционирования дозиметрического оборудования при проведении измерений в полях импульсного микросекундного фотонного излучения с известными характеристиками

This paper continues the series of articles [1,2,3] devoted to the determination of the main dosimetric characteristics of the fields of high-energy pulsed microsecond photon radiation generated by medical linear electron accelerator (LINAC for short) and assessment of the possibility of using LINAC as a source of reference radiation field for calibration of dosimetric equipment. This publication presents the results of evaluation of the portable dosimeter DKSAT1123 and detection unit BDKG-206 when measuring in such fields under different modes of LINAC operation. The possibility of this dosimetric equipment operation in the fields was assessed by comparing dosimeter readings with reference values of dosimetric characteristics of these fields, namely: average ambient dose equivalent rate H*(10), measured with the help of high-precision reference dosimeter DKS-AT5350. It was found that the deviation of H*(10) measurements conducted with the help of the DKS‑AT1123 and BDKG-206 from the reference values strongly depends on the radiation properties, namely the H*(10) value in the pulse. It is shown that at the H*(10) value in the pulse less than 0,2 mSv/s the deviation does not exceed 10%, at the H*(10) value in the pulse 0,2–0,4 mSv/s the deviation does not exceed 33%, and at the H*(10) value in the pulse more than 0,5 mSv/s the deviation can reach 72%.

Estimation of H*(10) and E(Ω) (and H*) using the Hp(10) and Hp(0.07) detectors of a body TLD

The H*(10) and E(Ω) responses to photons from 15 to 1253 keV were determined as a function of source orientation for the 7LiF:Mg,Cu,P ‘body’ and ‘skin’ elements of a thermoluminescence dosemeter optimized to measure Hp(10) and Hp(0.07), and are presented here relative to the results from 662 keV calibration exposures. The calculations were performed using the Monte Carlo transport code MCNP6.2, with the dosemeter located either free-in-vacuo or on a concrete wall; the impacts of the radiation field and dosemeter mounting are discussed. The relative response profiles were found to be fairly flat at energies above a few 10s of keV, leading to the conclusion that the personal dosemeter may be used in some environmental or workplace fields to provide reasonable estimates of ambient dose equivalent and effective dose; the latter is additionally relevant given the newly proposed operational quantity ambient dose (H*). Optimal correction factors were also derived that allow the dosemeter's standard Hp(d,0°,137Cs) system calibrations to be maintained.

Aircraft radiation exposure assessment in a radioactive environment: Ambient dose equivalents and effective doses

The release of radioactive plumes can occur due to nuclear power plant accidents, terrorist attacks, or scenarios in the chemical, biological, radiological, and nuclear defense (CBRN) field. Rescue aircraft may be exposed to ionizing radiation in such contaminated zones. Previous experimental assessments of these scenarios to estimate the potential dangers crew and passengers face are usually intangible. Using predictive software, it is possible to extract information from an external radiation field to which an aircraft would be subject. However, such codes cannot internally estimate the impact of this field on the aircraft. Using simulations based on the Monte Carlo method, it was possible to reproduce the scenario of plumes in the external environment (atmosphere) and, from this scenario, generate relevant information about their impact on the aircraft's internal environment. These assessments comprise the interaction of radiation from a radioactive plume with the aircraft structures and fuel tanks to estimate the radiation doses received by the crew and onboard electronics. This work evaluates the fluence rate, ambient dose equivalent rate (Ḣ*(10)), and effective dose rate (Ė) inside an aircraft flying through a radioactive plume resulting from a nuclear power plant accident. The simulation results suggest that radiation dose rates vary widely depending on the position within the aircraft. The ambient dose equivalent rate for photons varies by approximately 80% depending on the position within the aircraft. These differences reach around 98% for the ambient dose equivalent and effective dose rates for electrons. The data obtained may also be incorporated into risk assessments and support the development of protective measures to counter CBRN events.

Validation of ΓKa for diagnostic radiography in three different hospitals in the Lambayeque region, Peru

Air Kerma was measured and compared with calculated air Kerma aiming to validate the use of air Kerma gamma factors. The air Kerma was measured in radiology departments in three hospitals at the Lambayeque region in Peru. In each department were obtained three times different radiographs, for each radiograph the X-ray operational parameters were recorded such us: voltage, current, and triggering time; the average values of these parameters, and the measured air Kerma were calculated. In addition was recorded the focal spot-to r distance, and the filter thickness. Monte Carlo methods were used to calculate the spectra, air Kerma, and Ambient dose equivalent for an X-ray tube (W target, Al2O3 filter) working to 40, 50, and 60 kV; these values and those previously reported were used to calculate the spectra mean energy, the air Kerma and the Ambient dose equivalent gamma factors in function of voltage. Mean values of measured voltage were used to calculate the gamma factors for air kerma and this with the current, and triggering time mean values, in addition to the focal spot-to-r distance, and the filter thickness were used to calculate the air Kerma. The relative difference between calculated air Kerma and the measured air Kerma varies from −6.84 to 10.82% at the Belen hospital, −2.22 to 9.64% at Las Mercedes hospital and from 0.41% to 6.30% at the Ferreñafe hospital; thus, calculated air Kerma values using the gamma factors were validated with measured air Kerma values. Air Kerma gamma factors can be used for dosimetric purposes using the operational features of the X-ray tube with W target and Al2O3 filter; in addition can be used during the design of primary barriers of radiology hall.

Present and new operational quantities evaluated from photon spectrum measurements at workplaces in the research reactor and accelerator facility at the JAEA

The International Commission on Radiation Units and Measurements (ICRU) has proposed to change the definitions of the operational quantities used for the area and individual monitoring of external exposure in ICRU Report 95. Since introducing new operational quantities into the radiation monitoring may affect the dose assessment using the present personal dosimeters, precise assessments of the influence on monitoring are necessary by characterizing the energy spectrum in the workplace and the energy dependency of the dosimeters to be used. This study measured photon spectra using a NaI(Tl) scintillation detector or a LaBr3(Ce) scintillation detector at the Japan Research Reactor No. 3 (JRR-3) and Japan Proton Accelerator Research Complex (J-PARC) workplaces at the Japan Atomic Energy Agency (JAEA). The photon energy spectra were obtained by unfolding the pulse height distributions measured at each workplace. The present and new operational quantities were then evaluated and compared using the spectra measured at the workplaces.In JRR-3, the photon energy spectra were measured at the workplaces in the reactor room and thermal neutron beam hall while the reactor was operated. The ambient dose equivalent rates H˙*(10) at the workplaces were 0.2 μSv·h−1 – 13 μSv·h−1. High energy gamma-rays of 6–7 MeV from 16N were observed near the fuel failure detection system in the reactor room. Those of 7–8 MeV produced by the thermal neutron capture of iron were observed in the reactor room and the beam hall. In J-PARC, the photon energy spectra were measured at the workplaces in the beam tunnel of the 3 GeV synchrotron beam ring five days after the operation. The ambient dose equivalent rates H˙*(10) at the workplaces were 35 μSv·h−1 –430 μSv·h−1.It was found that the new personal doses, Hp, were 10%–20% smaller than the present personal dose equivalents, Hp(10), at the workplaces.

Design and testing of an irradiation room with low room-scattering for neutron calibration

IntroductionThis laboratory plans to establish neutron reference radiation fields with three neutron sources to calibrate neutron-measuring devices. To perform calibration at multiple dose rates, neutron ambient dose equivalent rate H˙*(10) needs to range 1 μSv/h to 10 mSv/h. The lower limit requires that the maximum available calibration distance should be at least 4.5 m. MethodsTo reduce room-scattered neutrons and extend the available calibration distance, MC simulations were conducted to determine the material and size of the irradiation room. A 3″ Bonner sphere and a LB6411 environmental neutron dosimeter were used to characterize the irradiation room. ResultsA 14.32 × 14.32 × 12.00 m3 irradiation room was built based on simulation results. Floor, roof, and walls are made of 75 cm concrete covered by a coating layer of 2 cm BPE and 3 cm PE. Experimental maximum available calibration distance reaches 4.65 m. The range of H˙*(10) for calibration covers 1 μSv/h to 10 mSv/h. Neutron and photon H˙*(10) outside the room are within 0.19 μSv/h and 0.22 μSv/h, respectively.

Statistical evaluation of individual external exposure dose of outdoor worker and ambient dose rate at evacuation ordered zones after the Fukushima Daiichi Nuclear Power Station accident

Following the accident at the Fukushima Daiichi Nuclear Power Station, evacuation orders were issued for the surrounding communities. In order to lift the evacuation order, it is necessary to determine individual external doses in the evacuated areas. The purpose of this study was to determine the quantitative relationship between individual external doses and ambient dose rates per hour as conversion coefficients. More specifically, individual external doses of Tokyo Electric Power Company Holdings employees in difficult-to-return zone were measured broadly over a long period (fiscal year 2020 to fiscal year 2022). To obtain highly accurate estimates, we used not only ambient dose rates based on airborne radiological monitoring data, but also Integrated dose rate map data that had been statistically corrected to correspond to local ambient dose rate gradients on the ground. As a result, the conversion coefficients based on the ambient dose rate map measured by airborne radiological monitoring were 0.42 for the Evacuation-Order Lifted Zones (ELZs), 0.37 for the Special Zones for Reconstruction and Rehabilitation (SZRRs), and 0.47 for the Difficult-to-Return Zones without SZRRs (DRZs). On the other hand, the conversion coefficients based on the Integrated dose rate map which is a highly accurate dose rate map based on statistical analysis of various types of monitoring that have been studied in government projects in recent years, were 0.78 for the ELZs, 0.72 for the SZRRs and 0.82 for the DRZs. Using these conversion coefficients, the individual external dose can be estimated from two representative ambient dose rate maps provided by the government.

A large and feasible national survey representative of population exposure to outdoor gamma radiation in urban areas.

Although data on outdoor gamma radiation are available for many countries, they have generally been obtained with measurements performed in undisturbed environments instead of in urban areas where most of the population lives. Only one large national survey, with on-site measurements in urban areas, has been identified worldwide, probably due to high costs (e.g., personnel and instrumentation) and difficulties in selecting measuring points. A campaign of outdoor gamma radiation measurements has been carried out in the entire Italian territory. All measurement points were selected at the infrastructures of an Italian telecommunications company as representatives of all the possible situations of outdoor exposure to gamma radiation for population in urban areas. Ten replicates of portable gamma (X) detectors carried out all the measurements. Approximately 4,000 measurements have been performed. They are distributed across 2,901 Italian municipalities, accounting for 75% of the Italian population. The national population-weighted mean of the gamma ambient dose equivalent rate (ADER) is 117 nSv h-1, and it ranges from 62 to 208 nSv h-1 and from 40 to 227 nSv h-1 for 21 regions and 107 provinces, respectively. The average variability at the municipal level, in terms of the coefficient of variation (CV) is 21%, ranging from 3 to 84%. The impact of land coverage and the distance from a building on the outdoor gamma radiation level was assessed with complementary measurements, leading to differences ranging from -40 to 50% and to 50%, respectively. A representative campaign of outdoor gamma dose rate measurements has been performed in Italy, only in urban areas, to assess the exposure effect due to outdoor gamma radiation on the population. It is the largest national campaign in urban areas worldwide, with a total of 3,876 on-site measurements. The land coverage and the distance from surrounding buildings were recognized to strongly affect outdoor gamma radiation levels, leading to high variability within small areas. The collaboration with a company that owns a network of facilities on a national territory as dense as the residing population made this survey feasible and affordable. Other countries might adopt this methodology to conduct national surveys in urban environments.