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Today, to respond to the increase of development of accurate, precise and relevant experiments in nuclear research reactors and tokamaks with their severe and intense operating conditions, there is a major need of innovative sensors that can accurately measure key parameters such as neutron and photon fluxes or nuclear heating rates. Thus, innovation in the field of nuclear instrumentation and measurements is a privileged research topic. It is crucial to develop optimized devices for accurate on-line in-core measurements, and scientific/technological requirements for various applications such as ageing of materials, safety applications, beam monitoring or nuclear physics. Nowadays, more and more semiconductors are used as sensor materials in nuclear instrumentation to measure various kinds of nuclear radiations. Silicon Carbide (SiC) is among them. In fact, SiC detectors could be used for the on-line measurement of key quantities such as neutron (thermal and fast) and photon fluxes. One main challenge is to enlarge the measurement range of this detector type. The work presented in this paper deals with this aim. Firstly, an introduction dedicated to the use of SiC versus other wide bandgap semiconductors and the characteristics of the studied sensor is shown. Secondly, this paper presents 3D numerical results obtained with a parametrical thermal study of the SiC detector using COMSOL Multiphysics code for a nuclear heating range corresponding to TRIGA nuclear research reactor conditions (integral neutron flux ~2.0 x1013 n·cm‑2·s-1 leading to a nuclear heating rate of 0.25 W·g-1 in Tungsten). The main objective is to adapt and optimize the design and the housing of the detector by determining its temperature field for different configurations. The influence of various parameters is presented such as that of the housing material nature, the gas nature around the diode, the gas-gap height and the housing thickness.

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were

We report on the low-energy electron and fast neutron irradiated 4H-SiC studied by deep-level transient spectroscopy (DLTS) and Laplace DLTS. Irradiations introduced two defects, E c −0.4 eV and E c −0.7 eV. They were previously assigned to carbon interstitial (Ci) labeled as EH1/3 and silicon-vacancy (V Si) labeled as S1/2, for the low-energy electron and fast neutron irradiation, respectively. This work demonstrates how Laplace DLTS can be used as a useful tool for distinguishing the EH1 and S1 defects. We show that EH1 consists of a single emission line arising from the Ci(h), while S1 has two emission lines arising from the V Si(h) and V Si(k) lattice sites.

Collection and interpolation of radiation observations is of vital importance to support routine operations in the nuclear sector globally, as well as for completing surveys during crisis response. To reduce exposure to ionizing radiation that human workers can be subjected to during such surveys, there is a strong desire to utilise robotic systems. Previous approaches to interpolate measurements taken from nuclear facilities to reconstruct radiological maps of an environment cannot be applied accurately to data collected from a robotic survey as they are unable to cope well with irregularly spaced, noisy, low count data. In this work, a novel approach to interpolating radiation measurements collected from a robot is proposed that overcomes the problems associated with sparse and noisy measurements. The proposed method integrates an appropriate kernel, benchmarked against the radiation transport code MCNP6, into the Gaussian Process Regression technique. The suitability of the proposed technique is demonstrated through its application to data collected from a bespoke robotic system used to conduct a survey of the Joz̆ef Stefan Institute TRIGA Mark II nuclear reactor during steady state operation, where it is shown to successfully reconstruct gamma dosimetry estimates in the reactor hall and aid in identifying sources of ionizing radiation.

A characterisation of cylindrical samples by Neutron Resonance Transmission Analysis (NRTA) at the GELINA facility of JRC Geel (Belgium) is presented. The samples were designed and produced for reactivity worth measurements in the MINERVE reactor of CEA Cadarache (France). NRTA was applied to determine the nuclide composition of UO2, Al2O3 and liquid samples that were doped with silver. The volume number densities of 238U, 107Ag and 109Ag obtained by NRTA are within 2 % fully consistent with the values that are quoted by the manufacturer. In addition, the NRTA data reveal a tungsten contamination which is not reported by the provider. It is shown that such a contamination contributes by up to 5.7 % to the reactivity worth.

The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly (SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offers a versatile and safe training and research facility since it can produce neutrons through nuclear fission reaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1 TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuel configurations. Based on extensive neutron transport simulations an SRA configuration is proposed, comprising 44 TRIGA fuel rods arranged in a 7 × 7 square lattice. This configuration is found to have a maximum k e f f value of 0.95001 ± 0.00009 at 4 cm pitch. The SRA is characterized by calculating the 3-dimensional neutron flux distribution and neutron spectrum. The effective delayed neutron fraction and mean neutron generation time of the system are calculated to be 748 pcm ± 7 pcm and 41 μs , respectively. Results obtained from this work will be the basis of the core design for the subcritical reactor facility that will be established in the Philippines.

With respect to the various European polices for water and environment protection, EU member states report to the European Commission through national monitoring programmes a wide variety of data on the pressures that aquatic ecosystems are facing. They further report data on the quality and quantity of water (“State of Environment”) at monitoring sites, and on the ecological and chemical status of surface waters under the Water Framework Directive. Many other databases on European waters (hydrographical databases as well as databases on state of water) were developed as a result of different European projects. There are also several sources about climatic data, data on river discharges, and data on pressures the water system is facing. All these databases are a welcome source of information, needed to analyse the pressures on and state of European inland waters, but they must be interlinked, harmonised, and entered into a common spatial database first. This was done within the MARS project, where we have interlinked and unified relevant available databases into a new spatial database. Data are now available in a new form to be used in analysis of state of inland aquatic ecosystems as a response to multiple pressures.

Introduction Potential changes in patient anatomy during proton radiotherapy may lead to a deviation of the delivered dose. A dose estimate can be computed through a deformable image registration (DIR) driven dose accumulation. The present study evaluates the accumulated dose uncertainties in a patient subject to an inadvertent breathing associated motion. Materials and methods A virtual lung tumour was inserted into a pair of single participant landmark annotated computed tomography images depicting opposite breathing phases, with the deep inspiration breath-hold the planning reference and the exhale the off-reference geometry. A novel Monte Carlo N-Particle, Version 6 (MCNP6) dose engine was developed, validated and used in treatment plan optimization. Three DIR methods were compared and used to transfer the exhale simulated dose to the reference geometry. Dose conformity and homogeneity measures from International Committee on Radioactivity Units and Measurements (ICRU) reports 78 and 83 were evaluated on simulated dose distributions registered with different DIR algorithms. Results The MCNP6 dose engine handled patient-like geometries in reasonable dose calculation times. All registration methods were able to align image associated landmarks to distances, comparable to voxel sizes. A moderate deterioration of ICRU measures was encountered in comparing doses in on and off-reference anatomy. There were statistically significant DIR driven differences in ICRU measures, particularly a 10% difference in the relative D98% for planning tumour volume and in the 3 mm/3% gamma passing rate. Conclusions T he dose accumulation over two anatomies resulted in a DIR driven uncertainty, important in reporting the associated ICRU measures for quality assurance.

An integrated modelling methodology for the calculation of realistic plasma neutron sources for the JET tokamak has been developed. The computational chain comprises TRANSP plasma transport and DRESS neutron spectrum calculations, and their coupling to the MCNP neutron transport code, bridging plasma physics and neutronics. In the paper we apply the developed methodology to the analysis of neutron emission properties of deuterium and helium plasmas at JET, and validate individual modelling steps against neutron diagnostic measurements. Two types of JET discharges are modelled-baseline-like and three-ion radio-frequency scenarios-due to their diversity in plasma heating, characteristics of the induced fast ion population, and the imprint of these on neutron emission properties. The neutron emission modelling results are quantitatively compared to the total neutron yield from fission chambers, neutron emissivity profiles from the neutron camera, neutron spectra from the time-of-flight spectrometer, and neutron activation measurements. The agreement between measured and calculated quantities is found to be satisfactory for all four diagnostic systems within the estimated experimental and computational uncertainties. Additionally, the effect of neutrons not originating from the dominating D(D, n)He-3 reactions is studied through modelling of triton burnup DT neutrons, and, in mixed D-He-3 plasmas, neutrons produced in the Be-9(D, n gamma)B-10 reaction on impurities. It is found that these reactions can contribute up to several percent to the total neutron yield and dominate the neutron activation of samples. The effect of MeV-range fast ions on the neutron activation of In-115 and Al-27 samples is measured and computationally validated.

A new approach for particle transport was used to calculate the Skyshine dose rates for the DEMO tokamak for the unsymmetrical case with one breeding blanket module removed for maintenance. Transport calculations for the large tokamaks, including DEMO and partly ITER, are usually performed in sector models postulating cyclic symmetry of geometry and source terms and applying reflecting boundary conditions on the sector edges. In the case of non-symmetric localized sources the sector models are not suitable. The calculation of shutdown dose rates, caused by one single tritium breeding blanket module moved from DEMO vacuum vessel into the upper maintenance hall, is an example of a localized radiation source and cannot be calculated in the standard way using a sector model. A new approach was followed for the calculations using the same sector model of DEMO, however with an according modification of the MCNP transport code. In this article the new approach is presented including its verification based on a simpler baseline model of DEMO. The results show that the obtained radiation dose distribution due to the one moved breeding blanket module differs by a factor of up to 2.8 from what would be obtained using a single sector model.

Fusion neutronics analysis before and after experiments at JET is traditionally performed using Monte Carlo particle transport code Monte Carlo N-Particle. For redundancy and diversity reasons there is a need of an additional Monte Carlo code, such as Serpent 2, capable of fusion neutronics analysis. In order to validate the Serpent code for fusion applications a detailed model of JET was used. Neutron fluxes and reaction rates were calculated and compared for positions outside the tokamak vacuum vessel, in the vacuum vessel above the plasma and next to a limiter inside the vacuum vessel. For all detector positions with DD and DT neutron sources the difference between neutron fluxes calculated with both Monte Carlo codes were within 2σ statistical uncertainty and for most positions (more than 90 % of all studied positions) even within 1σ uncertainty. Fusion neutronics analysis in the JET tokamak with Serpent took on average 10 % longer but this can be improved by changing the threshold value for determination of the transport method used. With the work presented in this paper the Serpent Monte Carlo code was validated to be a viable alternative to MCNP for fusion neutronics analysis for the JET tokamak.

In the last two decades we have assisted to a rush towards finding a 3He-replacing technology capable of detecting neutrons emitted from fissile isotopes. The demand stems from applications like nuclear war-head screening or preventing illicit traffic of radiological materials. Semiconductor detectors stand among the strongest contenders, particularly those based on materials possessing a wide band gap like silicon carbide (SiC). We review the workings of SiC-based neutron detectors, along with several issues related to material properties, device fabrication and testing. The paper summarizes the experimental and theoretical work carried out within the E-SiCure project (Engineering Silicon Carbide for Border and Port Security), co-funded by the NATO Science for Peace and Security Programme. The main goal was the development of technologies to support the fabrication of radiation-hard silicon carbide detectors of special nuclear materials. Among the achievements, we have the development of successful Schottky barrier based detectors and the identification of the main carrier life-time-limiting defects in the SiC active areas, either already present in pristine devices or introduced upon exposure to radiation fields. The physical processes involved in neutron detection are described. Material properties as well as issues related to epitaxial growth and device fabrication are addressed. The presence of defects in as-grown material, as well as those introduced by ionizing radiation are reported. We finally describe several experiments carried out at the Jozef Stefan Institute TRIGA Mark II reactor (Ljubljana, Slovenia), where a set of SiC-based neutron detectors were tested, some of which being equipped with a thermal neutron converter layer. We show that despite the existence of large room for improvement, Schottky barrier diodes based on state-of-the-art 4 H -SiC are closing the gap between gas- and semiconductor-based detectors regarding their sensitivity.

In this paper we present a computational exercise for shut-down dose rate calculations for the JET tokamak using the in-house developed JSIR2S code package as part of its validation. The computation is performed in two parts: neutron transport and transport of secondary gamma radiation. In order to calculate neutron activation reaction rates with sufficiently low variance, hybrid variance reduction techniques using the ADVANTG code have been utilized. Probability based sampling of secondary source particles was performed. Calculated gamma dose rates after shut down are compared with dose rate measurements performed on site using ionization chambers. The C/E agreement for 1st octant is between 0.8 to 1 while statistically meaningfull results for the 2nd octant are yet to be obtained.

Neutron detection is a relevant topic in the field of nuclear instrumentation. It is at the heart of the concerns for fusion applications (neutron diagnostics, measurements inside the Test Blanket Modules TBM) as well as for fission applications (in-core and ex-core monitoring, neutron mapping or safety applications in research reactors). Moreover, due to the even more harsh conditions of the future experimental reactors such as the Jules Horowitz Reactor (JHR) or International Thermonuclear Experimental Reactor (ITER), neutron detectors need to be adapted to high neutron and γ fluxes, high nuclear heating rates and high temperatures. Consequently, radiation and temperature hardened sensors with fast response, high energy resolution and stability in a mixed neutron and γ environment are required. All these requirements make wide-bandgap semiconductors and, more precisely, Silicon Carbide (SiC) serious candidates due to their intrinsic characteristics in such extreme environments. Thus, since the last decades, SiC-based detectors are developed and studied for neutron detection in various nuclear facilities. In this paper, a SiC-based neutron detector is 3-D designed and studied through thermal and radiation-matter interaction numerical simulations for a future irradiation campaign at the Jožef Stefan Institute TRIGA Mark II research reactor in Slovenia. Firstly, this paper presents the scientific background and issues of our SiC-based detectors. In a second part the 3-D geometry is shown. Thereafter, the 3-D numerical thermal simulation results are reported. Finally, the 3-D numerical radiation/matter interaction simulations results are presented.

The European Nuclear Experimental Educational Platform – ENEEP is currently being established by five European educational and research organizations in the framework of a Horizon 2020 project, initiated in 2019. The ENEEP partner institutions are the Jožef Stefan Institute (JSI, Slovenia), the Slovak University of Technology in Bratislava (STU, Slovak Republic), the Czech Technical University in Prague (CTU, Czech Republic), Technische Universität Wien (TU Wien, Austria) and the Budapest University of Technology and Economics (BME, Hungary). ENEEP is intended as an open educational platform, offering experimental hands-on education activities at the ENEEP partner facilities. ENEEP education activities will be offered in different formats (group and individual) and are targeted at university students at all educational levels and young professionals in the nuclear field. This paper gives an overview of the ENEEP project activities and the progress achieved thus far, highlighting the experimental capabilities which will be offered. In the first implementation phase, ENEEP will be based on a comprehensive set of experiments comprising the basics in Reactor Physics and Nuclear Engineering curricula, as well as more specific experiments focusing on particular aspects – investigated phenomena, types and working principles of detectors, etc. Subsequently, novel education activities will be introduced and implemented in ENEEP, following scientific development in nuclear science and technology and nuclear instrumentation detectors, stemming from research activities. Attention will be devoted to the development and optimization of remote education capabilities at the ENEEP partner institutions, of particular relevance during the current Covid-19 pandemic, which is responsible for major changes in education activities worldwide.

In 2016, the “E-SiCure” project (standing for “Engineering Silicon Carbide for Border and Port Security”), funded by the NATO Science for Peace and Security Programme was launched. The main objective is to combine theoretical, experimental and applied research towards the development of radiation-hard SiC-based detectors of special nuclear materials (SNM), with the end goal to enhance border and port security barriers. Prototype neutron detectors, configured as 4H-SiC-based Schottky barrier diodes, were developed for the detection of secondary charged particles (tritons, alphas and lithium atoms) which are the result of thermal neutron reactions on 10B and 6LiF layers above the surface of the 4H-SiC diodes. We designed a stand-alone prototype detection system, consisting of a preamplifier, shaping amplifier and a multichannel analyser operated by a laptop computer, for testing of neutron detector prototypes at the Jožef Stefan Institute (JSI) TRIGA reactor using a broad beam of reactor neutrons. The reverse bias for the detector diode and the power to electronic system were provided by a standalone battery-powered voltage source. The detector functionality was established through measurements using an 241Am alpha particle source. Two dedicated experimental campaigns were performed at the JSI TRIGA reactor. The registered pulse height spectra from the detectors, using both 10B and 6LiF neutron converting layers, clearly demonstrated the neutron detection abilities of the SiC detector prototypes. The computed neutron detection sensitivity of the single prototype detectors demonstrates that scaling SiC detectors into larger arrays, of dimensions relevant for border and port radiation detectors, could enable neutron sensitivity levels matching gas-based detector technology.

The COVID-19 induced restrictions have prevented reactor physics students from attending in-person reactor physics exercises which are a vital part of their education. Jožef Stefan Institute has organized remote exercises with the help of off-the-shelf technology, including multiple videoconferencing setups, remote desktop software, portable cameras, a dome camera, shared spreadsheets, and a common whiteboard. The students were encouraged to actively participate in the exercises by giving instructions to the reactor operator, asking and answering questions, logging data, operating digital acquisition systems, and performing analysis during the exercise. The first remote exercises were organized as a five-day course of experimental reactor physics for students from Uppsala University. The feedback was collected after the course using an anonymous online form and was generally positive but has revealed some problems with sound quality which were resolved later. The Jožef Stefan Institute can also organize a remote course during a full lockdown when the reactor is not able to operate using the in-house developed Research Reactor Simulator based on a point kinetics approximation and a simple thermohydraulic module.

The design and test of a robot-compatible, radiation detection instrument providing simultaneous $\gamma $ -ray imaging and $\gamma $ -ray spectroscopy is described. The sensing system comprises a cerium bromide inorganic scintillation detector and a cylindrical, lead slot collimator that is configured with a robot-compatible, on-board data acquisition system. The mount for the sensor is a lightweight, bespoke 3-axis gimbal actuated by servos for pan, tilt and rotation of the collimator for imaging capability. This paper discusses the integration of this relatively low-cost radiation detection apparatus with a commercially available, Robot-Operating-System-controlled robotic platform (a Clearpath Robotics™ Jackal). The detection system is compliant with the power and mass payload constraints of the robot. Its performance has been evaluated by means of two practical examples: a) measurements in a laboratory environment to assess the ability of the system to resolve two caesium-137 point sources, and b) deployment at the Jožef Stefan Institute TRIGA Mark II research reactor to assess the ability of the system to characterise the $\gamma $ -ray emission at 1 kW from a horizontal tangential beam port in the reactor hall and from the reactor sample pool above the core.

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, allows for real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and subcritical facilities. This paper presents a novel methodology for modeling control rods movement within the RAPID’s Fission Matrix (FM) formulation. This methodology, referred to as FM Control Rods “deltas” (FM-CRd), allows for combination and interpolation of RAPID’s FM coefficients to take into account the effect of inserting the control rods into the reactor core, while simultaneously limiting the amount of pre-calculations required. The FM-CRd methodology is benchmarked against experimental data from the JSI TRIGA ICSBEP benchmark and reference Serpent continuous-energy Monte Carlo calculations. The results show that RAPID’s FM-CRd can accurately calculate 3-D fission source distribution and k eff in the JSI TRIGA in a matter of seconds, hence constituting a powerful analysis tool for online analysis of the reactor during experiments.

The measurements of physical parameters of the TRIGA reactor and Nuclear power plant Krško (NEK) reactor cores have been in the past performed on hand of the neutron flux signal obtained from uncompensated ionization cells and by employment of the a digital meter of reactivity (DMR). At the TRIGA reactor only one ionization cell is currently used for flux measurements. During the insertion of one control rod the neutron flux distribution is significantly altered affecting the flux measurements of inserting different control rods. The problem is presently solved by assigning a correction factor to each control rod what introduces an additional uncertainty. In the present paper the implementation of four fission cells for reactivity measurements is presented. In this way determining the correct gamma background and its subtraction, performed by DMR algorithms, becomes less important as previously by using ionization chambers. The larger number of detectors also reduces the flux redistribution effects on the signal during individual control rod movements.

In this paper, the irradiation of glycerol and ethylene glycol by either mixed (neutron-γ) or γ-only (γ) fields at the TRIGA reactor of the Jožef Stefan Institute is described. This is highly relevant to future applications of fission reactor systems to produce useful feedstock derivatives from organic waste, beyond the production of heat and power. Samples of glycerol and ethyl glycol have been exposed to neutron-gamma radiation with fast neutron fluxes ranging from 7.7×1010 to 3×1012 cm−2s−1 and gamma-only irradiation at maximum dose-rates of 492 and 10 kGy hr−1, respectively. A study of the dependence of product yield versus absorbed dose has been conducted, for total dose ranges of 1 to 100 kGy. The products of these exposures have been identified through Gas Chromatography – Mass Spectrometry (GC-MS) techniques. Analysis comparing neutron-gamma irradiated samples of ethylene glycol and glycerol with gamma-irradiated samples shows no detectable qualitative difference between either irradiation type. Although, additional radiolysis products were detected when compared with available literature; ethyl acetate from ethylene glycol and solketal from glycerol. Quantitatively, neutron-gamma irradiation seems to be less effective at producing acetaldehyde from ethylene glycol, compared with gamma-only which can be explained through neutron moderation and consequent heating effects due to the borosilicate vials.

Deep level defects created by implantation of light-helium and medium heavy carbon ions in the single ion regime and neutron irradiation in n-type 4H-SiC are characterized by the DLTS technique. Two deep levels with energies 0.4 eV (EH1) and 0.7 eV (EH3) below the conduction band minimum are created in either ion implanted and neutron irradiated material beside carbon vacancies (Z1/2). In our study, we analyze components of EH1 and EH3 deep levels based on their concentration depth profiles, in addition to (&minus, 3/=) and (=/&minus, ) transition levels of silicon vacancy. A higher EH3 deep level concentration compared to the EH1 deep level concentration and a slight shift of the EH3 concentration depth profile to larger depths indicate that an additional deep level contributes to the DLTS signal of the EH3 deep level, most probably the defect complex involving interstitials. We report on the introduction of metastable M-center by light/medium heavy ion implantation and neutron irradiation, previously reported in cases of proton and electron irradiation. Contribution of M-center to the EH1 concentration profile is presented.

A pulse experiment model was validated in order to support future pulse experimental campaigns. All pulse experiments data was collected and are publicly available at http://trigapulse.ijs.si/. A comparison of the measured pulse physical parameters (maximal power, total released energy and full width at half maximum) and theoretical predictions (Fuchs-Hansen and the Nordheim-Fuchs models) was made.

The JSI TRIGA reactor features several in-core and ex-core irradiation facilities, each having different properties, such as neutron/gamma flux intensity, spectra and irradiation volume. A series of experiments and calculations was performed in order to characterise radiation fields in irradiation channel thus allowing users to perform irradiations in a well characterised environment. Since 2001 the reactor has been heavily used for radiation hardness studies for components used at accelerators such as the Large Hadron Collider (LHC) at CERN. Since 2010 it has been extensively used for testing of new detectors and innovative data acquisition systems and methods developed and used by the CEA. Recently, several campaigns were initiated to characterise the gamma field in the reactor and use the experimental data for improvement of the treatment of delayed gammas in Monte Carlo particle transport codes. In the future it is planned to extend the testing options by employing pulse mode operation, installation of a high energy gamma ray irradiation facility and allow irradiation of larger samples at elevated temperature.

An approach for calculating ex-core detector response using Monte Carlo code MCNP was developed. As a first step towards ex-core detector response prediction a detailed MCNP model of the reactor core was made. A script called McCord was developed as a link between deterministic program package CORD-2 and Monte Carlo code MCNP. It automatically generates an MCNP input from the CORD-2 data. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krško nuclea power plant. For the hot zero power configuration, the deviations are within 3 % for majority of fuel assemblies and slightly higher for fuel assemblies located at the core periphery. The computational model was further verified by comparing the calculated control rod worth to the CORD-2 results. The deviations were within 50 pcm and considered acceptable. The research will in future be supplemented with the in-core and ex-core detector signal calculations and neutron transport outside the reactor core.

In 2016, the NATO Science for Peace and Security Programme funded research project ”Engineering Silicon Carbide for Border and Port Security” — E-SiCure was launched, its objective being the development of radiation-hard silicon carbide (SiC) based detectors of special nuclear materials (SNM), with the aim to enhance border and port security barriers. Detector prototypes based on SiC Schottky Barrier Diodes (SBDs) and neutron converter films were developed. This paper presents the results of a dedicated experimental testing campaign performed at the Jožef Stefan Institute (JSI) TRIGA reactor in which several SiC detector prototypes equipped with 10 B and 6 LiF converter films were irradiated in the Dry Chamber of the reactor. The obtained results demonstrate a clearly measurable neutron response, which varies linearly with the neutron flux. The measured particle spectra from the SiC detectors exhibit a clear structure, attributable to the nature and energy of secondary particles originating as reaction products from nuclear reactions involving 10 B and 6 Li isotopes. The determined sensitivity of the detectors, their active volume being 1 mm × 1 mm × 25 μ m , 1 mm × 1 mm × 69 μ m and 1 mm × 1 mm × 170 μ m , was of the order of 2 × 10−5 counts per second, per unit of neutron flux [counts s−1 per n cm−2s−1] (for neutron energies between 0 and 5 eV). Scaling the detection sensitivity by a factor of 1 0 5 , i.e. to an array with a surface of around 20 cm × 2 m, comparable to large B F 3 or 3 He detectors, would theoretically enable an overall sensitivity of around 2 counts s−1 per n cm−2s−1, which is already comparable to typical neutron sensitivity values of gas detectors, in the range from several to over 100 counts s−1 per n cm−2s−1. Due to its outstanding tolerance to harsh environments (including high temperatures and radiation fields) and superior electronic properties when compared to other semiconductors, SiC is a promising base material for the fabrication of solid-state detectors with stable and long life-time. Improvements in sensitivity combined with the capability of fabricating large modules (SiC arrays), could make SiC an important detection technology, applicable also in the context of border and port security barrier monitoring.

In this article we present a first part of the results obtained during an irradiation campaign conducted at the Jozef Stefan Institute to observe the behaviour of piezoelectric materials under gamma and neutron flux. Specific instrumentation has been developed and has enabled the monitoring throughout the irradiation of several materials such as lead zirconate titanate (PZT) or modified Bismuth Titanate (BiT) in either massive or thick film form. Various parameters such as resonance frequency, electromechanical coupling coefficient, electrical capacitance, dielectric losses were measured as a function of the flow and dose received. The results obtained confirm that the samples work up to doses of 10 18 n°/cm2 and that the behaviour of the samples varies according to their composition and form.

International audience; The neutron field of various irradiation positions of the TRIGA Mark II reactor of the Jožef Stefan Institute (JSI) has been thoroughly characterized by neutron activation dosimetry and miniature fission chambers techniques. In order to have a fully validated calculation scheme to analyze and plan experiments, the gamma field also has to be experimentally validated. The 10-year long collaboration between CEA and JSI is a perfect framework to carry out such a study, and measurements of the gamma field started in late 2016. Several measurement techniques were investigated in in-core and ex-core positions. On-line measurements were carried out using miniature ionization chambers manufactured by the CEA and PTW Farmer. Positional dependence was studied, showing a decrease in the delayed gamma contribution to the total gamma flux with increasing distance from the reactor core center. To characterize the gamma dose in the core, as well as in the periphery, thermo-and optically-stimulated luminescent detectors (TLDs and OSLDs, respectively) were tested. These detectors are commonly used at CEA to measure the gamma dose in a given material, in order to study the nuclear heating in various core elements (control rod, baffle, structural material). Different filters were used in order to assess an integrated dose ranging from a few Gy up to several kGy. Comparisons of experimental results against calculations performed with the JSIR2S code package show a very good agreement. The feasibility of such measurements demonstrates the complementarity between measurements with dosimetry and ionization chambers from low to very high gamma-dose environment, such as in material testing reactors.

Fast neutrons propagate significantly in free space with interaction properties that contrast markedly with those of electromagnetic radiation , the latter being fundamental to most wireless communications . Here, we report on the configuration and operation of a nuclear instrumentation set-up designed to transmit digitally encoded information using fast neutrons. We demonstrate the potential of fast neutron radiation as a medium for wireless communications for applications where conventional electromagnetic transmission is either not feasible or is inherently limited.