Gamma Radiation Field Research Articles

Technical basis for extended dry storage of aluminum-clad spent nuclear fuel

This paper presents the technical basis for extended dry storage of aluminum-clad spent nuclear fuel (ASNF). This basis addresses and closes five knowledge gaps defined by the U.S. Department of Energy (DOE) Spent Nuclear Fuel Working Group (SNFWG) in DOE-ID/RPT-1575 “Aluminum-clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 Years) Dry Storage.” The SNFWG's primary concern pertaining to dry storage of ASNF is rooted in the presence of (oxy)hydroxide layers adherent to the aluminum cladding of the spent fuel that formed in service, indicating a risk of radiolytic molecular hydrogen (H2) generation in a strong gamma radiation field—a process that can be associated with disintegration of (oxy)hydroxide and the molecular decomposition of chemisorbed and physisorbed water bonded to it. Potential consequences are overpressurization of sealed ASNF dry storage canisters and creation of a flammable atmosphere in sealed and vented canisters. Further, there is the concern that the generation of other chemical species due to chemical reactions, such as nitric acid (HNO3), could lead to canister component corrosion. To address these issues, experimental work was conducted, including investigations of the performance of surrogate aluminum specimens and actual ASNF material featuring adherent (oxy)hydroxide layers during wet or dry storage. Further, a comprehensive characterization of the sample material was completed. Drying experiments were executed, providing better estimations of the amount of residual, chemisorbed, and physisorbed water after ASNF dry storage canister conditioning. These empirical activities allowed for the development of predictive numerical models to simulate the chemical species yield within sealed and vented dry storage canisters for 50 years. The findings of these activities lead to the main conclusion that extending current ASNF dry storage durations by 50 years is both safe and viable.

Low-Dose Gamma Radiation Fields Decrease Cell Viability, Damage DNA, and Increase the Expression of Hsp70 and p53 Proteins in Human Leukocytes

Low-Dose Gamma Radiation Fields Decrease Cell Viability, Damage DNA, and Increase the Expression of Hsp70 and p53 Proteins in Human Leukocytes

The Influence of Neutron and Gamma Radiation Fields on Gas-Filled Surge Arresters

Abstract The study examines the effect of neutron and gamma radiation on commercial gas surge arresters. The research is of experimental-theoretical nature. The experimental part of the research was performed un-der well-controlled laboratory conditions. The combined measurement uncertainty was about 5%. The experimental system is specially designed for the observed problem and has certain original solutions. The test procedure was fully automated and had software support in the experiment management and also for data collection and statistical processing. The obtained results show that neutron and gamma ra-diation improves the functional characteristics of gas surge arresters with a memory effect. The obtained results are explained in accordance with the theory of interaction of neutron and gamma fields with ma-terial and with the theory of electric discharge in gases. The results presented in this study are important for the design of surge protection in systems that can be found in the field of neutron and gamma radia-tion because they can achieve a positive synergistic effect of protection in hybrid schemes with other components for surge protection whose characteristics spoil this radiation.

Development of a Novel MCNP-OSCAR Fusion Interface for the 3-D Assessment of Gamma Dose Due to the Activated Corrosion Products

The gamma radiation field due to the activated corrosion products (ACPs) represents a key input for shielding design, licensing, and planning of maintenance operation. The present work aims to define and implement a reliable calculation chain based on the Outil de Simulation de la ContAmination en Réacteur–tOol of Simulation of ContAmination in Reactor (OSCAR)-Fusion and Monte Carlo N-Particle Transport (MCNP) codes for the assessment of 3-D gamma dose rate due to the ACPs. In particular, a piece of software has been developed to make easy the exchange of information between the codes: the MCNP neutron spectrum is processed to calculate the OSCAR neutron reaction rates; based on the OSCAR output and geometrical data, the 3-D gamma source is produced and exported in the surface source write/surface source read (WSSA/RSSA) MCNP format. Finally, a pure MCNP gamma transport provides the 3-D gamma dose rate distribution.

Investigation of the light output of 3D-printed plastic scintillators for dosimetry applications

Three-dimensional (3D) printing, specifically digital light processing (DLP) technique, can be used to manufacture plastic scintillators of any shape. The purpose of this study was to determine the light output of DLP 3D-printed scintillators for dosimetry applications. Two types of plastic scintillators with dimensions 10 mm × 10 mm × 10 mm were fabricated using DLP 3D-printing at Hanyang University, South Korea. The light output of these DLP 3D-printed samples was measured and compared to that of a commercial plastic scintillator of the same dimensions, RP-408, produced by casting. The 3D-printed scintillators emitting violet and blue light had a lower relative light output by 49% and 43%, respectively, compared to the RP-408 reference scintillator. We also investigated three types of scintillator surface finishing methods: the original surface made by the 3D printer, a sanded surface, and a polished surface. Furthermore, three wrapping configurations were tested: bare scintillator, diffuse-type polytetrafluoroethylene tape, and specular-type enhanced specular reflector foil. Both reflector types, diffuse and specular, reflected blue light with comparable efficiency. Additionally, emission and transmission spectra of the samples were measured. Emission maxima were located at 430 nm for RP-408, and 438 and 475 nm for two 3D-printed samples. Transmittance at the wavelength of maximum emission was equal to 89% for RP-408, and 73% and 66% for the two DLP-printed samples. Although the light output of the 3D-printed scintillators was about 50% lower than that of the commercial plastic scintillator, due to characteristics of 3D-printed plastic scintillators, i.e. fast, low-cost production, and easy customization of the printed shape, they are promising as an active part of dosimeters for use in high intensity gamma radiation fields produced by medical linear accelerators with acceptable signal-to-noise ratio level.

Study of certain congenital malformations due to low-level radiation exposures from high background radiation areas

The coastal regions of some states (Kerala, Tamil Nadu, Andhra Pradesh and Orissa) in India are known to have plenty of thorium-rich minerals, therefore possess high external gamma radiation fields, and are identified as Natural High Background Radiation Areas. There are concerns about congenital malformations due to the high background radiation prevailing in the region.Therefore, scientists find interest to study the possible biological effects of low-level exposures on the local populace who are living with generations under these extraordinary radiation fields. The present study is an attempt to find the correlation between the incidences of certain congenital malformations through a 1:3 matched case-control study. A total of 75 subjects (58 mental retardation cases and 17 Cleft lip/palate cases) along with 225 controls were selected for the investigations. Indoor and outdoor gamma dose rates were measured using a halogen-quenched GM tube-based survey meter, while the indoor inhalation dose rates were measured based on the radon and progeny measurements using the twin cup dosimeters equipped with LR-115 Type II solid-state nuclear track detectors. Conditional logistic regression analysis was used to find out the correlation between the radiation dose with congenital malformation. This study found an insignificant risk of cleft lip/palate (but not for mental retardation) from the exposures to natural high external gamma radiation fields in the study area. Therefore, the results of the investigation indicate no explicit risk of congenital malformations by the high background natural radiation prevailing in the region.

3-D gamma radiation field reconstruction method using limited measurements for multiple radioactive sources

The reconstruction of 3-D gamma radiation field is important for radiation protection in the presence of unknown radiation sources and limited measurements, such as the decontamination and decommissioning of nuclear facilities. While there may be multiple radioactive sources in these applications, the current 3-D reconstruction methods do not take the multiple radiation sources into consideration. This paper develops a method that has the potential to reconstruct the 3-D gamma radiation field of multiple radioactive sources with limited measurements of the field. This method solves a 3-D fourth-order TV-H-1 equation to promote the continuity of the reconstructed 3-D radiation field, which simultaneously fills the missing dose rates and recovers local details near the radioactive sources. The proposed method was verified against three Monte Carlo simulations with different numbers of sources and validated against a 1-D experiment. The method was tested with different numbers of combinations of the input data, including both randomly and regularly distributed data sets. The results reveal that the method accurately reconstructs the 3-D radiation field and achieves a relative error level as low as 2.18%, using only 3.125% of the data for the simulation cases. The 1-D experiment shows a relative error range between 5.57% and 65.28%. These results preliminary demonstrate the potential of the proposed method for 3-D radiation field reconstruction in practice.

Determination of double-layer gamma build-up factor using Monte Carlo code, FLUKA: Development of new empirical formula

Designing shield for coupled neutron and gamma radiation field is a complex problem as it involves combination of high-Z and low-Z materials in composite or laminar form with optimized thickness. In the present work we have utilized general purpose open-source Monte Carlo code, FLUKA for estimation of gamma exposure build-up factors (EBF) of single and multilayer shields. Excellent agreement of FLUKA calculated results of EBF for iron, water and poly-methyl methacrylate (PMMA) (for different energy point-isotropic gamma source) with standard codes/database like EGS4 and ANSI, demonstrates the validity and reliability of this code for the present purpose. Subsequently, the code is utilized for determination of double-layer exposure build-up factor (DEBF) of iron/PMMA shield. In this regard, a new empirical formula is constructed for predicting DEBF of any arbitrary combination of layer materials, applicable over wide range of layer thicknesses and gamma energies. The coefficients of thus constructed empirical formula have been associated with radiation interaction parameters of individual components. Further, the proposed formula is shown to predict DEBF of iron/water shield with good accuracy, thereby ascertaining the robustness of the formula. The quantitative accuracy with which the proposed formula is able to predict DEBF for both Iron/PMMA and Iron/Water shield lies within 12% of FLUKA calculated data.

Route measurements of natural surface radiation background in the Almaty region

Measurements of the spatial distribution of radon isotopes were carried out from April 2021 to August 2021 in the foothills of the Trans-Ili Alatau of the Tien Shan in the Almaty region at various heights above sea level: from 500 to 2500 meters. They were carried out using electronic radiometric equipment: beta-dosimeter “RKS-01B-SOLO”; gamma dosimeter “RKS-01G-SOLO”; radiometer of radon and its daughter decay products “RAMON- 02” in the field. As a result, preliminary assessment schemes were built for route measurements of the 222Rn radon isotope, beta and gamma radiation fields from natural daughter products of decay of radon isotopes and radionuclides located in the surface atmospheric layer.

Estimated measurements of natural background radiation in the surface atmospheric layer of the Almaty region

The radon isotopes of natural radioactive radiation of the earth’s crust emanating to the surface areconcentrated in the human environment and have a direct impact on the people health. Due to this factthe problem of studying the radiation distributions in the surface atmospheric layer of the Earth is urgent.In this work, measurements of the exposure dose rate, beta particle flux, and equivalent equilibriumvolumetric activity (EEVA) of radon in the surface atmospheric layer of the Earth have been performed.Measurements of radon and gamma, beta backgrounds were carried out from April 2021 to August 2021in the Almaty region at various altitudes above sea level: from 500 to 2500 meters. They were carried outusing electronic radiometric equipment: beta-dosimeter “RKS-01B-SOLO”; gamma dosimeter “RKS-01GSOLO”;radiometer of radon-222 “RAMON-02” in the field. As a result, preliminary assessment schemeswere built for route measurements of the 222Rn radon isotope, beta and gamma radiation fields fromnatural daughter products of decay of radon isotopes and radionuclides located in the surface atmosphericlayer. As a result of this work, an altitude dependence of changes in the natural radiation gamma and betabackground of the surface surface layer is presented, which grows with an increase in the surface above sealevel. The different activity of radon over rocks and soils is shown, which, as is known, is associated withthe exhalation of radon and which depends, among other things, on the density of the soils. The preliminaryresults of measurements of significant perturbations of the EEVA of radon did not reveal. Key words: variations in time of radon; natural beta active radionuclides; coagulation; atmospheric radioactive, nanoparticles.

Characterization and thermoluminescence study of gamma irradiated Tb-doped ZnO and undoped ZnO synthesized by spray pyrolysis method

High gamma dose-resistant undoped ZnO and Tb-doped ZnO thermoluminescent (TL) micro-phosphors were prepared by the spray pyrolysis method. Scanning electron microscopy shows crystalline rods with hexagonal morphology, (0.1-0.4 µm diameter, and about 1 µm length). Raman spectra dispersion reveals a würtzite form. Photoluminescence (PL) study of irradiated zinc oxide films indicates the generation of defects produced by gamma irradiation resulting in an increased probability of electron-hole exciton recombination. PL spectrum shows emission bands from 5D4-7Fj=6,5,4,3 transitions ascribed to Tb3+ dopant in zinc oxide phosphor. X-ray diffraction patterns for both types of films growth (undoped ZnO and Tb-doped ZnO) are typical of zinc oxide crystalline structure, with no noticeable effect of Tb ions. Dosimetric properties, for both samples, show a low TL fading signal and TL reproducibility signal for undoped ZnO and Tb-doped ZnO samples was 29 and 57 %, respectively. The kinetic parameters such as activation energy E, frequency factor s, and Rm values, were obtained by Computerized Glow Curve Deconvolution (CGCD) assuming Mixed Order Kinetic model (MOK). The results show that the MOK well described the glow curves of zinc oxide films. The heating rate effects produced a broadening of glow peak located at 420 K. For purposes like radiation detector, atomic effective number (Zeff) was obtained: 27.74 and 56.47 for undoped ZnO and Tb-doped ZnO samples, respectively. The samples were exposed to gamma radiation in a wide range of 0.25–20 kGy dose. TL properties of undoped ZnO and Tb-doped ZnO samples show that these materials could be used to detect high doses in a gamma radiation field.

Evaluation of the corrosion behavior of modern spent nuclear fuels under repository conditions

Abstract. In Germany it is planned to directly dispose spent nuclear fuel (SNF) from nuclear power plants together with other high-level radioactive wastes (HLW) from former SNF reprocessing (e.g., vitrified waste), in a deep geological repository for heat-generating wastes – the siting process for this repository was started in 2017 and is ongoing. Based on several decades of research, development, and demonstration (RD&D) it is generally accepted at the technical and scientific level that direct disposal of HLW and SNF in deep mined geological repositories is the safest and most sustainable option (CEC, 2011; IAEA, 2004). The current efforts to improve the performance and accident tolerance of fuels in nuclear power generation resulted in an increased utilization of a variety of new types of light-water reactor (LWR) fuels such as fuels doped with Cr, Al, and Si. This doping leads to a significant change of the microstructure of the fuel matrix. The corrosion behavior of these types of fuels under conditions relevant to deep geological disposal has hardly been studied so far; however, this is of crucial importance as the development of a robust safety case for deep geological disposal of SNF requires a solid understanding of its dissolution behavior over very long time scales (up to 1 million years). To fill this knowledge gap, additional systematic studies on modern doped UO2 fuels were needed. Corrosion experiments with SNF cannot entirely unravel all of the various concurring effects of the dissolution mechanism due to the chemical and structural complexity of SNF and its high beta and gamma radiation field during the first 1000 years; moreover, technical restrictions only allow a very limited number of experiments. Therefore, within the EU-DisCo project (https://www.disco-h2020.eu, last access: 11 October 2021), a very ambitious programme of corrosion studies on irradiated Cr and Al/Cr doped fuels was carried out, which was complemented by systematic single-effect dissolution studies (e.g., with respect to doping level, grain size and thermodynamic aspects) performed on carefully prepared and characterized, simplified UO2-based model materials. Here, we present recent results on the dissolution behavior of tailor-made UO2 model materials in accelerated static batch experiments using H2O2 as simulant for radiolytic oxidants, present in long-term disposal scenarios for SNF in failed container conditions due to the alpha irradiation of water. In these dissolution experiments pure UO2 reference pellets exhibiting different densities and grain sizes, as well as Cr-doped UO2 pellets with various Cr-doping levels, produced using different doping methods having different grain sizes, were used. In addition, Nd-doped and industrially produced Cr- and Cr/Nd-doped UO2 pellets were used to determine the influence of these parameters on the dissolution rates. The dissolution experiments were performed under strictly controlled conditions with respect to exclusion of oxygen, temperature control, and exclusion of light. This bottom-up approach was followed to understand how the addition of Cr-oxide into the fuel matrix affects SNF dissolution behavior under repository relevant conditions. The results of the dissolution experiments performed with real SNF and the model materials obtained by the DisCo partners build the basis for numerical simulations on the dissolution behavior of modern SNF. First results of the data evaluation indicate that the addition of dopants and the consequential modification of the fuel matrix does not lead to a significant change of the dissolution behavior of these fuels under repository relevant conditions compared to standard SNF (i.e. dissolution rates agree within an order of magnitude).

Comparison of detectors with pulse shape discrimination capability for simultaneous detection of gamma-rays, slow and fast neutrons

In this paper, five detectors (EJ-270 — single plastic scintillator, EJ-309B5 — single liquid scintillator, and 3 phoswich detectors: EJ-276+GS20, GS20+EJ-276, EJ-309+GS20) were studied in terms of simultaneous detection of gamma-rays and slow and fast neutrons. The measurements were carried out in a mixed field of neutron and gamma radiation from an intense (∼106 neutrons/s/4π) PuBe source. The PuBe neutron source was placed in an 18 cm diameter paraffin ball in order to increase the number of slow neutron detections. This paper covers the measurements of the neutron/gamma discrimination, photoelectron yield and analysis of the light pulse shapes originating from events related to gamma-rays, slow neutrons and fast neutrons. Sensitivity to slow neutrons is also briefly discussed. A comparison of n/γ discrimination for all of the tested detectors shows excellent performance of all phoswich detectors. Phoswich configurations tested within the scope of this study allow unambiguously resolve all three types of radiation and show superior performance in comparison with single loaded plastic or liquid scintillators.

New approach to the conceptual design of STUMM: A module dedicated to the monitoring of neutron and gamma radiation fields generated in IFMIF-DONES

International Fusion Materials Irradiation Facility — DEMOOriented Neutron Source (IFMIF-DONES) is a planned powerful neutron source, which will generate an intense flux of neutrons (up to ∼1015n/s/cm2) with a fusion-relevant energy spectrum. It will be an accelerator source based on deuteron beam - lithium target reactions. The engineering design of IFMIF-DONES is elaborated in the frame of the Early Neutron Source work package of the EUROfusion consortium. The facility will be dedicated to the irradiation of suitable structural materials planned for the construction of future fusion reactors such as DEMO (Demonstration Fusion Power Plant). Start-up Monitoring Module (STUMM) is designed to monitor radiation and thermal conditions during the commissioning phase of IFMIF-DONES, characterize the produced neutron flux and validate neutronic modeling of the facility. The conceptual design of STUMM is prepared by a team of physicists and engineers from the Institute of Nuclear Physics Polish Academy of Sciences (IFJ PAN) and the National Centre for Nuclear Research (NCBJ), Poland. This paper presents the concept of STUMM, the proposed design of the module, and selected measuring systems.

Characterization of a polyphenyl ether oil irradiated at high doses in a TRIGA Mark II nuclear reactor

Samples of a commercially available polyphenyl ether (PPE) oil are irradiated in the mixed neutron and gamma radiation field of the Pavia TRIGA Mark II reactor, at irradiation doses ranging from 0.78 MGy to 7.8 MGy. The oil is a nuclear grade product, whose radiation resistance reported in the product specification is based on cobalt-60 gamma irradiation data. The collection of additional data in a nuclear reactor radiation field performed in the present work, is motivated by the need to use this oil in neutron-rich environments such as targets and accelerators.The post-irradiation characterization of the oil samples here reported includes viscosity measurements, total acid number (TAN), Fourier transform infrared spectroscopy (FT-IR), gas chromatography (GC), gel permeation chromatography (GPC) and nuclear magnetic resonance (NMR) spectroscopy.The presented results evidence a remarkable increase of kinetic viscosity with the absorbed dose. Differences in the radiation-induced viscosity increase are observed comparing the data collected in mixed neutron and gamma radiation with the previous ones collected in cobalt-60 gamma radiation. This result emphasizes the need of further studies on the dependence of the radiation damage on different radiation environments. A negligible increase of TAN is reported. The FT-IR, GC, GPC and NMR show progressive polymerization with increasing dose. The increasing molecular size is in line with the viscosity data presented.

DEVELOPMENT OF A MODELING APPROACH TO ESTIMATE RADIATION FROM A SPENT FUEL ROD QUIVER

Before encapsulation of spent nuclear fuel in a geological repository, the fuels need to be verified for safeguards purposes. This requirement applies to all spent fuel assemblies, including those with properties or designs that are especially challenging to verify. One such example are quivers, a new type of containers used to hold damaged spent fuel rods. After placing damaged rods inside the quivers, they are sealed with a thick lid and the water is removed. The lid is thick enough to significantly reduce the amount of the gamma radiation penetrating through it, which can make safeguards verification from the top using gamma techniques difficult. Considering that the number of quivers at storage facilities is foreseen to increase in near future, studying the feasibility of verification is timely. In this paper we make a feasibility study related to safeguards verification of quivers, aimed at investigating the gamma and neutron radiation field around a quiver designed by Westinghouse AB and filled with PWR fuel rods irradiated at the Swedish Ringhals site. A simplified geometry of the quiver and the detailed operational history of each rod are provided by Westinghouse and the reactor operator, respectively. The nuclide inventory of the rods placed in the quiver and the emission source terms are calculated with ORIGEN-ARP. The radiation transport is modeled with the Serpent2 Monte Carlo code. The first objective is to assess the capability of the spent fuel attribute tester (SFAT) to verify the content for nuclear safeguards purposes. The results show that the thick quiver lid attenuates the gamma radiation, thereby making gamma radiation based verification from above the quiver difficult.

Response of 4H-SiC Detectors to Ionizing Particles

We report the response of newly designed 4H-SiC Schottky barrier diode (SBD) detector prototype to alpha and gamma radiation. We studied detectors of three different active area sizes (1 × 1, 2 × 2 and 3 × 3 mm2), while all detectors had the same 4H-SiC epi-layer thickness of approximately µm, sufficient to stop alpha particles up to 6.8 MeV, which have been used in this study. The detector response to the various alpha emitters in the 3.27 MeV to 8.79 MeV energy range clearly demonstrates the excellent linear response to alpha emissions of the detectors with the increasing active area. The detector response in gamma radiation field of Co-60 and Cs-137 sources showed a linear response to air kerma and to different air kerma rates as well, up to 4.49 Gy/h. The detector response is not in saturation for the dose rates lower than 15.3 mGy/min and that its measuring range for gamma radiation with energies of 662 keV, 1.17 MeV and 1.33 MeV is from 0.5 mGy/h–917 mGy/h. No changes to electrical properties of pristine and tested 4H-SiC SBD detectors, supported by a negligible change in carbon vacancy defect density and no creation of other deep levels, demonstrates the radiation hardness of these 4H-SiC detectors.

Linearity and rate capability measurements of RPC with semi-insulating crystalline electrodes operating in avalanche mode

The intrinsic rate capability and the ageing properties of the Resistive Plate Chambers are closely related to the electrodes material and to the front-end electronics threshold. The development of a low noise pre-amplifier led us to improve the intrinsic rate capability of High Pressure Laminate (bakelite) RPCs up to ∼10 kHz/cm2, nevertheless the effective rate is significantly limited by electrodes ageing. To further improve the effective rate capability new materials are investigated. A Resistive Plate Chamber with crystalline semi-insulating Gallium Arsenide electrodes has been characterized with high energy electrons beam at the Beam Test Facility (BTF), (INFN National Laboratory of Frascati, Italy). The response of the Resistive Plate Chamber to multiple bunched electrons was measured operating the detector in avalanche mode. The intrinsic rate capability has been also measured operating the detector in a uniform high energy gamma radiation field at the GIF++ facility (EHN1 of SPS, CERN).

GET 8–2019 State Primary Standard of Units of Air Kerma, Air Kerma Rate, Exposure, Exposure Rate, and Energy Flux of X-Radiation and Gamma Radiation

Results of studies designed to improve GET 8–2011, the State Primary Standard of the units of air kerma, air kerma rate, exposure, exposure rate, and energy flux of x-radiation and gamma radiation, are described. The studies are performed in 2019 to implement the recommendations of Report No. 90, Key Data for Dosimetry of Ionizing Radiation: Standards and Applications, International Commission on Radiation Units and Measurements. The following changes are introduced into the equation for reproduction of the units of quantities in GET 8–2019: in the field of x-radiation, new correction factors for free-air ionization chambers are introduced and the value of the relative standard uncertainty of the mean ionization energy is changed; in the field of gamma radiation, the values of the product of the mean ionization energy and ratio of the mass stopping powers of graphite to air in measurements with cavity ionization chambers are changed. More accurate values of the units of quantities reproduced by GET 8–2019 are obtained.

Ambient radioactivity and atmospheric electric field: A joint study in an urban environment

Ambient radioactivity and atmospheric electricity are inextricably linked phenomena. In order to assess the role of ambient radioactivity in the local variability of the atmospheric electric field at an urban site, simultaneous measurements of radon concentration, gamma radiation, and atmospheric electric field are carried out in the city of Porto, Portugal. Both radon and gamma radiation display an average daily cycle peaking before sunrise, but with considerable variability from day to day, particularly in amplitude. The atmospheric electric field displays a daily cycle with a minimum at dawn and maximum in the early afternoon, as well as a secondary peak in the early morning. The temporal variation of the daily patterns is analysed by means of an empirical orthogonal function analysis, and related to local meteorological parameters. The variability of the local atmospheric electric field is mainly determined by aerosol transport and accumulation close to the surface associated with local meteorological conditions and atmospheric stability rather than by conductivity variations associated with ambient radioactivity.