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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

In this paper, detailed verification and experimental validation of the formulations and algorithms of the Multi-stage Response-function Transport (MRT)–based Real-time Analysis for Particle-transp...

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.

Neutron transmission and capture measurements have been performed at the time-of-flight (TOF) facility GELINA of the EC-JRC-Geel, using metallic discs of natural silver with different thicknesses. Resonances of neutron interactions with 107Ag and 109Ag were analysed in the energy region below 1 keV applying the Reich-Moore approximation of the R-Matrix theory. Discrepancies between the experimental data and calculations based on the neutron resonance parameters recommended in the main neutron cross section libraries (JEFF, ENDF/B, JENDL) were observed. Improved resonance parameters were determined by a resonance shape analysis using the REFIT code. The resonance energy, neutron (Γn) and radiation (Γγ) width were adjusted in a simultaneous fit to transmission and capture yield spectra below 100 eV. Above 100 eV fixed average radiation widths ( Γ γ 107 =140 meV, Γ γ 109 = 130 meV) were used and the neutron width Γn was adjusted to only the capture yield data. Capture resonance integral I0 were calculated from the parameters derived in this work (I0,107 = 101.20 b, I0,109 = 1515.24 b) and compared with those calculated using the cross sections recommended in the main libraries. Discrepancies in I0 for 107Ag are mainly due to difference in the parameters of the resonances at 16.4 eV, 41.6 eV, 173.9 eV and 202.8 eV. In case of 109Ag differences are mainly due to the difference in parameters of the resonance at 5.2 eV.

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.

Well characterized neutron and gamma fields inside a nuclear reactor are of key importance for its safe operation and for successful utilization of various research reactor irradiation facilities. In case of high-flux research reactors such as BR2 in Belgium, Maria in Poland and the future Jules Horowitz Reactor in France, the gamma energy deposition rate in reactor structural components and irradiated samples reaches values well over 10 W/g. To assure safe reactor operation, the gamma field and associated heating must therefore be thoroughly characterized in order to provide adequate component and sample cooling. The gamma field can be divided into two contributions: prompt gamma rays are emitted almost instantly after neutron interaction with nuclei, while the delayed gamma rays are emitted from nuclei, which become radioactive by neutron absorption, generated from fission and other processes. Most modern Monte-Carlo particle transport codes enable the transport of prompt gamma rays; a few support delayed gamma ray generation and transport as well. The latter have mostly been applied to fusion devices, where detailed shutdown dose-rate measurements have been performed. Although the delayed gamma field can also be simulated in fission devices, significant inaccuracy in the result is to be expected due to the computational complexity arising from the large number of radioactive fission products and incompleteness of nuclear data. Furthermore, the unavailability of experimental delayed gamma measurements in fission systems presents an important challenge for the validation of the experimental results. Previous measurements in several research reactors show that the delayed gamma flux amounts to around 30 % of the total gamma flux. However, these evaluations were performed with measurement data obtained during rapid reactor shutdowns (SCRAMs), using a single measurement point per SCRAM. In this paper we propose a new technique to accurately determine the magnitude of the delayed gamma component and its time evolution, based on synchronous acquisition of fission and ionization chamber signals. The measurements were performed at the JSI TRIGA reactor, using fission and ionization chambers placed in several in-core measurement positions. Their signal was acquired synchronously and at the highest possible acquisition rate in order to distinguish between measurement noise and reactor transients. Using the novel delayed gamma extraction technique we were able to estimate the magnitude of the delayed gamma contribution to be: 18.9 % ± 2.0 % at the reactor core periphery, linearly increasing towards the reactor core center to 31.4 % ± 2.8 % of the total measured gamma flux signal after 10 min of reactor operation.

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.

Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate neutron and prompt gamma particles emitted from TRIGA research reactor during operation. Firstly, the method was validated by measuring dose rates around open beam port number 5 was unplugged. Neutron and gamma dose rates inside the reactor hall in the vicinity of the beam port were calculated and compared to the measurements. Due to the satisfactory agreement, the method was later used to design external shielding for the same beam port when it was upgraded - special mechanism was installed that allows irradiation of larger samples. Computational analysis of the proposed shielding configuration provided acceptable dose rate levels inside the reactor hall. When the shield was constructed, calculated dose rates were confirmed by the actual measurements. No modifications were needed.

The spatial reference framework is the lateral extent of the river channel and its floodplain, named “flood-prone area extent”. Due to human interventions into the hydrological cycle and morphological alterations of the river channel and its floodplain, some areas that were regularly flooded once, may not experience such flooding today. We characterize them as “former floodplains”. Floodplains flooded now are named “active floodplains”. The analysis is done on the spatial resolution level named “Functional Elementary Catchment” (FEC) of the European catchments and Rivers network system (Ecrins) database and for the flood-prone areas in Europe that include former and active floodplains with river channels. It is named “Potential flood prone area”. In the first step we defined floodplains typology. For the assessment part we developed indicators of floodplain forms and processes, defined their benchmark condition and performed quality classification. Here, we describe what spatial data we used and what data we still miss to produce reliable assessment. The spatial layer “Potential flood-prone area extent” was derived from two spatial layers, Potential Riparian Zone Delineation of the Copernicus Land Monitoring Service and JRC flood hazard map for Europe 100-year return period, a result of flood model “LisFlood”. The candidate list of typology factors included 31 factors derived from various databases such are Ecrins, MARS, FAO, Copernicus, WorldClim, PCGLOBEWB and IHME. Factors represent abiotic state before human intervention into rivers and floodplains and are grouped into regions, climate, morphology, hydrology, geology and physics – river dynamics. The calculated factors are reasonably covering the assessment area (95% - 99%) with the exception of the physics – river dynamics factors. This information was obtained for less than 30% of European area. The selection of factors defining floodplain types was based on the criterion of adequate spatial coverage, reliability and non-redundancy. As a result, floodplain types were derived from seven factors, three morphological (river average altitude and slope and average floodplain width), one geological (dominant catchment geo-chemistry) and three hydrological factors (specific runoff as mean annual discharge divided by catchment area, high flow duration and high flow pulses). Hydrological and morphological factors are only approximations to the natural state, so we propose to further develop databases providing information on river and floodplain hydromorphology prior to major human interventions.Indicators of floodplain forms are derived from two layers, Riparian Zone Land Cover/Land Use and High Resolution Water & Wetness of the Copernicus Land Monitoring Service. The land use layer provides a good basis for assessing the current distribution of floodplain habitats. We also estimate the size of the active natural floodplains using wetness data, but the results can be improved with systematic European wide information on present hydrotechnical structures and hydromorphological alterations. Such data would also support assessment of floodplain ecological condition and management options.

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.

Developing new state-of-the-art, low-cost and radiation hard detectors is an extremely difficult challenge which can be tackled only by a multisciplinary group of scientists and engineers from various fields having access to different infrastructure.