Mixed Neutron Research Articles

Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten

Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten

Implementation of optically simulated luminescent dosimeter for quality control of gamma ray dose of an accelerator-based neutron source.

Neutron beams utilized for performing BNCT are composed of a mixture of neutrons and gamma rays. Although much of the dose delivered to the cancer cells comes from the high LET particles produced by the boron neutron capture reaction, the dose delivered to the healthy tissues from unwanted gamma rays cannot be ignored. With the increase in the number of accelerators for BNCT, a detector system that is capable of measuring gamma ray dose in a mixed neutron/gamma irradiation field is crucial. Currently, BeO TLDs encased in quartz glass are used to measure gamma ray dose in a BNCT irradiation field. However, this type of TLD is no longer commercially available. A replacement dosimetry system is required to perform the recommended ongoing quality assurance of gamma ray measurement for a clinical BNCT system. The purpose of this study is to evaluate the characteristics of a BeO OSLD detector system under a mixed neutron and gamma ray irradiation field and to assess the suitability of the system for routine quality assurance measurements of an accelerator-based BNCT facility. The myOSLD system by RadPro International GmbH was evaluated using the accelerator-based neutron source designed for clinical BNCT (NeuCure BNCT system). The readout constancy, linearity, dose rate effect, and fading effect of the OSLD were evaluated. Free-in-air and water phantom measurements were performed and compared with the TLD results and Monte Carlo simulation results. The PHITS Monte Carlo code was used for this study. The readout constancy was found to be stable over a month-long period and similar to the TLD results. The OSLD readout signal was found to be linear, with a high coefficient of determination (R2≥0.999) up to a proton charge of 3.6C. There was no significant signal fading or dose rate dependency. The central axis depth dose and off-axis dose profile measurements agreed with both the TLD and Monte Carlo simulation results, within one standard deviation. The myOSLD system was characterized using an accelerator system designed for clinical BNCT. The experimental measurements confirmed the OSLD achieved similar, if not superior to, the currently utilized dosimetry system for routine QA of an accelerator-based BNCT system. The OSLD system would be a suitable replacement for the current TLD system for performing routine QA of gamma ray dose measurement in a BNCT irradiation field.

Performance assessment and further optimization of the activation based criticality dosimetry system at the Belgian nuclear research centre SCK CEN

In places where fissile materials are being handled, the risk of a criticality accident can never be absolutely eliminated. Criticality accidents require specialized dosimetry techniques, which differ markedly from those used in routine radiological protection. This is caused by the fact that one should be able to measure very high doses up to 10 Gy in mixed gamma and neutron fields with very high instantaneous dose rates up to 105 Gy/s and complex energy spectra. The criticality dosimetry system at the Belgian Nuclear Research Centre SCK CEN is accredited according to ISO 17025 by the Belgian accreditation body BELAC and is used to monitor about 4000 nuclear workers. The system is based on a dedicated criticality dosemeter with four activation detectors (gold foils with and without cadmium shielding, an indium foil and sulphur pellets) and an iterative unfolding algorithm starting from a guess neutron fluence energy spectrum. This work describes the design of the SCK CEN criticality dosemeters and the iterative unfolding algorithm. Furthermore, it also discusses two new developments. Firstly, effective neutron reaction cross sections were calculated by MCNP6.2 Monte Carlo radiation transport simulations to be used in the unfolding algorithm. These effective cross sections take into account the effects of the dosemeter, the potential presence of the body and the neutron angle of incidence. These effects were demonstrated to be important, especially for the gold foils and if the dosemeter is worn against the body. Secondly, a Bayesian unfolding algorithm based on a parametric description of the neutron fluence energy spectrum was developed and tested. The results of this new algorithm were in good agreement with the iterative algorithm. The uncertainty assessment was more realistic with the Bayesian algorithm because it includes also the uncertainty associated with the unfolding process. However, for spectra that cannot be well described by the parametric function, the Bayesian algorithm can lead to significant under- or overestimations. In case sufficient information is available, it is better to make use of the iterative algorithm with a realistic guess spectrum to start from. Finally, this work also demonstrated the compliance of the SCK CEN criticality dosimetry system with the IAEA criteria for criticality dosimetry systems.

Pencil beam kernel-based dose calculations on CT data for a mixed neutron-gamma fission field applying tissue correction factors.

Objective.For fast neutron therapy with mixed neutron and gamma radiation at the fission neutron therapy facility MEDAPP at the research reactor FRM II in Garching, no clinical dose calculation software was available in the past. Here, we present a customized solution for research purposes to overcome this lack of three-dimensional dose calculation.Approach.The applied dose calculation method is based on two sets of decomposed pencil beam kernels for neutron and gamma radiation. The decomposition was performed using measured output factors and simulated depth dose curves and beam profiles in water as reference medium. While measurements were performed by applying the two-chamber dosimetry method, simulated data was generated using the Monte Carlo code MCNP. For the calculation of neutron dose deposition on CT data, tissue-specific correction factors were generated for soft tissue, bone, and lung tissue for the MEDAPP neutron spectrum. The pencil beam calculations were evaluated with reference to Monte Carlo calculations regarding accuracy and time efficiency.Main results.In water, dose distributions calculated using the pencil beam approach reproduced the input from Monte Carlo simulations. For heterogeneous media, an assessment of the tissue-specific correction factors with reference to Monte Carlo simulations for different tissue configurations showed promising results. Especially for scenarios where no lung tissue is present, the dose calculation could be highly improved by the applied correction method.Significance.With the presented approach, time-efficient dose calculations on CT data and treatment plan evaluations for research purposes are now available for MEDAPP.

Mechanical and electrical performance of glass fiber reinforced plastic insulation for cryogenic application in fusion magnet irradiated in fast breeder reactor

In a superconducting (SC) magnet fusion reactor, the insulation material used has to withstand significant fluxes of neutrons and neutron-induced gamma radiation. Epoxy-based glass fiber-reinforced plastic (GFRP) composites are the insulation system for fusion magnets and are highly demanding. This insulation system requires excellent mechanical and electrical performance under mixed neutron and gamma radiation at cryogenic temperatures. The insulation material developed consists of boron-free S-glass fiber impregnated with a two-component modified Diglycidyl ether of Bisphenol-A (DGEBA) epoxy resin. An anhydride-cured DGEBA resin system is not suitable for cryogenic temperatures. Therefore, to overcome this thermal stress issue in cryogenic components of SC magnets, a high-toughness, flexible, aromatic amine-cured epoxy resin system was developed for the insulation system. In order to assess the radiation resistance for mechanical and electrical properties, the composite samples were irradiated in the fast breeder test reactor (FBTR) at IGCAR, Kalpakkam, India. The irradiation experiment was carried out for a neutron fluence of 1.0E+21 n/m2 and a gamma dose of 0.4 MGy at a full reactor power of 32 MWt. The tensile strength, interlaminar shear strength, and breakdown strength measurements were investigated before and after irradiation at room temperature and 77 K. The test results indicate that no significant degradation was observed in the properties of the irradiated samples up to the above-mentioned radiation doses.

Transition Metal Borides for All-in-One Radiation Shielding.

All-in-one radiation shielding is an emerging concept in developing new-generation radiation protection materials since various forms of ionizing radiation, such as neutrons and gamma rays, can occur simultaneously. In this study, we examine the ability of transition metal borides to attenuate both photon and particle radiation. Specifically, fourteen different transition metal borides (including inner transition metal borides) are selected for examination based on their thermodynamic stabilities, molecular weights, and neutron capture cross-sections of the elements they contain. Radiation shielding characteristics of the transition metal borides are computationally investigated using Phy-X/PSD, EpiXS and NGCal software. The gamma-ray shielding capabilities of the transition metal borides are evaluated in terms of the mass attenuation coefficient (μm), the linear attenuation coefficient (µ), the effective atomic number (Zeff), the half-value layer (HVL), the tenth-value layer (TVL), and the mean free path (MFP). The mass and linear attenuation factors are identified for thermal and fast neutrons at energies of 0.025 eV and 4 MeV, respectively. Moreover, the fast neutron removal cross-sections (∑R) of the transition metal borides are calculated to assess their neutron shielding abilities. The results revealed that borides of transition metals with a high atomic number, such as Re, W, and Ta, possess outstanding gamma shielding performance. At 4 MeV photon energy, the half-value layers of ReB2 and WB2 compounds were found as 1.38 cm and 1.43 cm, respectively. Most notably, these HVL values are lower than the HVL value of toxic Pb (1.45 cm at 4 MeV), which is one of the conventional radiation shielding materials. On the other hand, SmB6 and DyB6 demonstrated exceptional neutron attenuation for thermal and fast neutrons due to the high neutron capture cross-sections of Sm, Dy, and B. The outcomes of this study reveal that transition metal borides can be suitable candidates for shielding against mixed neutron and gamma radiation.

Modelling the response of CLLBC(Ce) and TLYC(Ce) SiPM-based radiation detectors in mixed radiation fields with Geant4

CLLBC(Ce) and TLYC(Ce) are novel scintillation materials capable of measuring mixed gamma ray and neutron radiation fields that have gained significant interest in the areas of space and nuclear safety/security science. To date Geant4, the world’s most popular Monte Carlo radiation modelling toolkit, has yet to be effectively used to simulate the full response of these materials when coupled to near ultra-violet Silicon PhotoMultipliers (SiPMs). In this work an experimentally validated Geant4 application has been developed with optimised material composition, optical data tables, and physics transport settings that is able to accurately simulate the response of CLLBC(Ce) and TLYC(Ce) SiPM-based radiation detectors under both gamma ray and neutron irradiation. Experimental benchmarking for five different radioactive sources (Co-60, Cs-137, Eu-152, Am-241, and Cf-252) illustrated that this developed Geant4 application was able to reproduce the position and structure of all major spectral features (full energy gamma ray photo-/neutron capture peaks, X-ray escape photopeaks, Compton edge, Compton backscatter peaks, and Compton plateau) to high level of accuracy.

“Radiation to Materials” at CERN

Irradiation tests are continuously needed at the European Organization for Nuclear Research (CERN), to experimentally assess the radiation tolerance of commercial non-metallic materials for use in high-radiation areas. The ’Radiation to Materials’ R2M activity organizes such tests and provides expertise in the interpretation of results, for the design and upgrade of devices absorbing critical doses during their lifetime. The various activities covered by the R2M are presented. Up to ten irradiation tests per year are organized in external <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">60</sup> Co facilities, in which samples absorb up to several MGy of gamma dose at dose rates in the kGy/h range. Cables, O-rings, lubricants, glues, optical materials, and insulators are examples of possible tested materials. In parallel, selected samples are irradiated in a mixed neutron and gamma irradiation station built close to the Neutron Time-Of-Flight facility. Dedicated research studies, mainly focused on oil, greases and elastomeric O-rings, are being performed to deepen the understanding of radiation effects in these critical materials. As an example, the mechanical and structural characterization of two elastomeric EPDM-based commercial materials irradiated in gamma radiation up to 3 MGy of dose is presented.

Biomarker integration for improved biodosimetry of mixed neutron + photon exposures

There is a persistent risk of a large-scale malicious or accidental exposure to ionizing radiation that may affect a large number of people. Exposure will consist of both a photon and neutron component, which will vary in magnitude between individuals and is likely to have profound impacts on radiation-induced diseases. To mitigate these potential disasters, there exists a need for novel biodosimetry approaches that can estimate the radiation dose absorbed by each person based on biofluid samples, and predict delayed effects. Integration of several radiation-responsive biomarker types (transcripts, metabolites, blood cell counts) by machine learning (ML) can improve biodosimetry. Here we integrated data from mice exposed to various neutron + photon mixtures, total 3 Gy dose, using multiple ML algorithms to select the strongest biomarker combinations and reconstruct radiation exposure magnitude and composition. We obtained promising results, such as receiver operating characteristic curve area of 0.904 (95% CI: 0.821, 0.969) for classifying samples exposed to ≥ 10% neutrons vs. < 10% neutrons, and R2 of 0.964 for reconstructing photon-equivalent dose (weighted by neutron relative biological effectiveness) for neutron + photon mixtures. These findings demonstrate the potential of combining various -omic biomarkers for novel biodosimetry.

Neutron spectrometry and dosimetry using a multi-shell Single Sphere Neutron Spectrometer with thermo-luminescent and optically stimulated luminescent detectors

A portable neutron spectrometer with concentric multi-shell moderating assembly and TL, OSL detectors has been validated with standard neutron reference fields to generate the neutron spectra and estimate the ambient dose equivalents, H*(10). Further, the system has been used in the mixed neutron–gamma radiation fields at accelerator facility. The neutron fluence and the ambient dose equivalent estimates were verified using the standard reference neutron fields, viz. 241Am-Be, 252Cf, D-T neutron sources. The system was then employed in the accelerator radiation environment to measure the neutron spectral yield from the interactions of 15 and 20 MeV protons with a thick natural Ta target. The results were compared with those obtained from FLUKA simulations. A conventional neutron dose equivalent meter was also used to verify the ambient dose equivalent estimated from the spectrometric measurements. The reproducibility of the neutron spectra and dose values were found to be satisfactory, thereby encouraging the implementation of the portable single sphere neutron spectrometer in complex radiation environment around particle accelerators. Present studies demonstrate the measurements with the in-house built spectrometer for neutron energies up to 20 MeV.

Americium-beryllium shield container optimization with new high alloyed stainless steel and composite material using MCNPX code

In this paper the goal is to optimize a shield for the americium-beryllium radioactive source, which generates a mixed neutron and gamma-ray radiation field, using two different materials designed with simultaneous moderation and absorption of neutrons and gamma-ray shielding in mind: a new type of stainless steel and epoxy resin (C21H25ClO5) composite with a 40% NiO additive. For this purpose, MCNPX was used and the equivalent dose rate for both types of radiation was calculated. After the optimization process, the shield volume experienced a 54% reduction while its weight, due to the use of a steel alloy, was increased by 3.56 %. At a distance of 1 m from the center of the source, the equivalent dose rate was reduced by 10.98% in comparison to that of the original system.

Dual-Particle Imaging Performance of a Cs2LiYCl6:Ce (CLYC)-Based Rotational Modulation Collimator (RMC) System

A dual-particle imager based on a rotational modulation collimator (RMC) utilizing a Cs <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sub> LiYCl <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">6</sub> :Ce (CLYC) scintillator was developed and investigated for its dual-particle imaging performance. Design considerations and previous simulation work were verified by conducting experiments, and experimental results showing neutron and gamma-ray imaging capabilities were obtained. The imaging performance of the CLYC-based RMC system was further evaluated by investigating it in a mixed neutron and gamma-ray environment, and dual-particle imaging capability of the RMC system was demonstrated. Signal-to-noise ratio (SNR) values larger than 45 were achieved for reconstructed images of each neutron and gamma-ray source distribution, if modulation patterns were obtained with average fractional counting uncertainty of less than 3%. The structural similarity (SSIM) index of reconstructed images was calculated to be almost unity, indicating an artifact-free image.

Novel metallic Bi-Pb–Cd-Ag alloys for shielding against neutrons and gamma rays

Obtaining high-performance shielding alloys against mixed neutrons and γ-rays is vital in the field of nuclear radiation. In this work, shielding Bi-Pb–Cd-Ag alloys with superior thermal neutron shielding and high ability to attenuate γ-rays were developed. The microstructure, melting behavior, and mechanical properties of the prepared alloys were characterized. The γ-rays mass attenuation coefficients () were obtained using WinXCom program in the energy range of 0.015–15 MeV. The shielding parameters including atomic and electronic cross-sections (σ a and σ e), effective atomic number (Z eff ), effective electron density (N eff ) were calculated. The exposure buildup factor (EBF) using a geometric progression (GP) fitting method was estimated. The experimental measurements were carried out at 662 and 1173, 1332 keV of 137Cs and 60Co isotope sources, respectively using a scintillation gamma spectrometer NaI(Tl) detector. The calculated results of WinXCom are consistent and in agreement with the experimental values of The macroscopic effective removal cross-section () of the samples for fast neutrons were calculated. Also, the attenuation of thermal neutrons through the prepared alloys has been investigated. It was found that the alloy containing 30% Cd complies with the compromise between superior thermal neutron attenuation besides its superior ability to attenuate γ-rays with good mechanical strength and high ductility.

Diagnostic of Fusion Neutrons on EAST Tokamak Using 4H-SiC Detector

Due to significant radiation hardness, thermal properties, low sensitivity to gamma rays, and fast response, the 4H-SiC detector exhibits promising properties in measuring the neutron production in the high-power fusion devices. To demonstrate the potential value of the 4H-SiC detectors as a neutron flux monitor and a neutron spectrometer on experimentally advanced superconducting tokamak (EAST) during the 2021 EAST deuterium–deuterium plasma experimental campaigns, the 4H-SiC detector covered with a 25- <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\mu \text{m}$ </tex-math></inline-formula> -thick <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">6</sup> LiF film 95% enriched in <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">6</sup> Li and the 4H-SiC detector covered with a 1.5-mm-thick polyethylene layer were installed on a viewport located at the mid-plane of port F outside the EAST vacuum vessel and operated continuously during the whole experimental campaign with the goal to measure both the total and the time-dependent neutron emission from the plasma. The results are compared with those obtained with detectors routinely working at EAST ( <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">235</sup> U Fission Chambers). Despite their small active volumes, the 4H-SiC detectors were able to measure the total and the time-dependent neutron emission with good reliability and stability. The results of this work not only demonstrated the satisfactory performance of the 4H-SiC detector in the harsh environment of a mixed neutron–gamma high-radiation field and strong electromagnetic field, but also its applicability as a neutron spectrometer and a neutron flux monitor for fusion deuterium–plasma experiments.

Simulation of Ionizing/Displacement Synergistic Effects on NPN Bipolar Transistors Irradiated by Mixed Neutrons and Gamma Rays

Transistors working in complex radiation environments such as space are simultaneously irradiated by neutrons and gamma rays. But the mechanism of the synergistic radiation effect between the two rays is still unclear. Based on TCAD, the synergistic radiation effects of ionizing/displacement damage caused by mixed neutrons and gamma rays are simulated. The results demonstrate that the synergistic effects are more serious than the simple sum of the two radiation effects due to their mutual enhancement. The change of the carrier recombination rate in the device at different positions shows that the displacement effects increase the peak value of surface recombination rate; meanwhile, the ionizing dose effects enhance the recombination process in bulk silicon. The mechanism of this phenomenon is that positive charges from the oxide layer and interface enhance the recombination of carriers in bulk, and the reduced carrier lifetime caused by defects from bulk makes carriers more likely to be trapped by the interface traps. In addition, the simulation result which shows the influence of temperature on the synergistic effects indicates that the synergistic effects are more sensitive to the lower temperature.

Synergistic effects in MOS capacitors with an Au/HfO2–SiO2/Si structure irradiated with neutron and gamma ray

The properties of oxide trapped charges and interface state density in the metal oxide semiconductor (MOS) capacitors with an Au/HfO2–SiO2/Si structure were investigated under irradiation of 14 MeV neutron and 60Co gamma-ray. In the mixed neutron and gamma irradiation environment, the formation of the oxide trapped charges in the HfO2–SiO2 layer is determined by the total deposited ionization energy, i.e. the sum of ionization energy deposition of the neutrons and the accompanying gamma rays, while the influence of the displacement damage caused by 14 MeV neutrons can be ignored. The interface state density depends not only on the ionizing energy loss (IEL) but also the non-IEL (NIEL), and NIEL plays a major role below the critical neutron fluence of 4.5 × 1012 n cm−2. The synergistic effect of the interface state is observed increases with energy deposition in the oxide at lower fluences, while decreasing above the critical fluence. These results confirm the existence of the synergistic effect of neutron and gamma irradiation in damaging HfO2 MOS devices.

Lead borate@polydopamine core–shell particles chemically bonded with silicone rubber for neutron and γ‐rays shielding

AbstractRubber composites with excellent radiation shielding and flexibility are extremely important to personal protective equipments (PPEs) for protecting workers from radiation hazards, especially for mixed radiations. It is, however, challenging to achieve uniform dispersion of fillers and good compatibility of the interfaces in highly‐filled rubber composites that are closely related to their physical properties. In this article, lead borate@polydopamine (PBO@PDA) core–shell particles are chemically bonded with silicone rubber (SR) for co‐shielding of neutron and γ‐rays. Uniform dispersion of the core–shell particles and good compatibility of the interfaces give rise to enhanced flexibility of the rubber composites. Particularly, the SR composite with 40 wt% PBO particles displays increases of 106% in elongation at break and 490% in tensile strength to neat SR. Furthermore, the mass attenuation coefficient of γ‐rays (105 KeV) reaches 2.35 and the thermal neutron absorption rate (0.025 eV) of is 76.9%. This work takes into account the balance between radiation shielding and flexibility of rubber composites, which provides a facile strategy to fabricate excellent integrated properties of flexible materials for shielding mixed neutron and γ‐rays.

Use of the chemical Fricke dosimeter and its modifications for dosimetry of gamma neutron radiation of a pulsed reactor

The paper investigates the characteristics of the chemical Fricke dosimeter (with the standard composition (D1), without NaCl addition to the solution (D2), without NaCl but with a tenfold increased concentration of Fe2+ (D3)) under continuous and pulsed irradiation with an ultra-high dose rate of the BARS-6 reactor with unshielded metallic cores. The dosimeter radiosensitivity had a linear dependence on the gamma neutron radiation dose in a range of 25 to 750 Gy and was respectively 1.96 ± 0.05 μGy–1 (D1), 2.04 ± 0.05 μGy–1 (D2), and 2.08 ± 0.5 μGy–1 (D3) in the continuous irradiation mode, and 1.24 ± 0.05 μGy–1, 2.00 ± 0.05 μGy–1, and 1.94 ± 0.05 μGy–1 in the pulsed irradiation mode. This makes ≈ 60% of their sensitivity to the 60Со gamma radiation (3.40 ± 0.02 μGy–1), and 36%, 1.6 times as less, for a standard Fricke dosimeter irradiated in the pulsed mode. The experimental value of the radiation chemical yield, Gn(Fe3+), for all solution modifications and both irradiation modes varied slightly and was 0.84 ± 0.11 μM/J on the average, except for the standard solution in the pulsed mode (0.66 ± 0.07 μM/J). The neutron doses determined by chemical and activation dosimeters coincided within the error limits, but the chemical dosimeter readings were systematically higher, by about 20%. Therefore, in the fission spectrum neutron dose rate range of 0.4 to 7×108 Gy/min, there is no dose rate effect both in the standard Fricke dosimeter version (without NaCl) and in the modified version, which makes it possible to use modified Fricke dosimeters to assess the physical and dosimetry characteristics of mixed gamma neutron radiation beams.

Isomeric ratios in neutron capture reaction, induced by thermal, resonant and mixed thermal-resonant neutrons on 114Cd and 116Cd nuclei

The isomeric ratios in neutron capture reactions, induced by thermal, resonant and mixed thermal-resonant neutrons on 114Cd and 116Cd nuclei have been measured by the activation method using off-line gamma spectroscopy. The data analysis and necessary corrections as the self-absorption and summing cascade effects were made in gamma rays radioactivity measurements to improve the precision of the isomeric ratio determination. The results were discussed and compared with those in the existing literature, which can contribute as additional data to the nuclear data banks and used for theoretical calculations in nuclear reaction models as well as for nuclear applications.

Low dose gamma irradiation pretreatment modulates the sensitivity of CNS to subsequent mixed gamma and neutron irradiation of the mouse head

ABSTRACТ Purpose: To explore if the total body γ-irradiation at a dose of 0.1 Gy 7 days prior to acute mixed γ, n-irradiation of the head at the dose of 1 Gy can reduce the harmful effects of neutron irradiation on the hippocampal functions, neuroinflammation and neurogenesis. Materials and methods: Mice were exposed to γ-radiation alone, mixed γ,n-radiation or combined γ-rays and γ,n-radiation 7 days after γ-irradiation. Two months post-irradiation, mice were tested in Open Field and in the Morris water maze. The content of microglia, astrocytes, proliferating cells and cytokines TGF-β, TNF-α, IL-1β, GFAP levels, hippocampal BDNF, NT-3, NT-4, NGF mRNA expression were evaluated. Results: Two months after combined irradiation, we observed impaired hippocampus-dependent cognition, which was not detected in mice exposed to γ,n-irradiation. Combined exposure and γ,n-irradiation led to a significant increase in the level of activated microglia and astrocytes in the brains. The level of pro- and anti-inflammatory cytokines in the brain and hippocampal neurotrophine's genes changed differenly after the combined exposure and γ,n-irradiation. The quantity of DCX-positive cells was reduced after γ,n-irradiation exposer alone, but increased after combined irradiation. Conclusions: Our results indicate radio-adaptive responses in brains of mice that were exposed to low-dose gamma irradiation 7 days prior to acute 1 Gy γ,n-irradiation.