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Following an initial Round Robin inter-comparison of gamma spectrometry measurements reported in 2014, this paper presents results from the first part of a second Round Robin inter-comparison commissioned by the European Working Group on Reactor Dosimetry in 2018. Measurements were performed by thirteen European organisations on a set of irradiated neutron activation detectors representative of those commonly used by the Reactor Dosimetry community to measure neutron fluence using gamma spectrometry methods. The radionuclides measured were 110mAg, 58Co, 60Co, 54Mn, and 46Sc. The purpose of the exercise was to demonstrate the level of consistency between participating organisations in blind tests of measurements. The samples used were disks of iron, nickel, titanium, and two standard alloys of aluminium, one of 1% cobalt and the other 1% silver. They were irradiated in the MARIA reactor operated by National Centre for Nuclear Research, Poland. Participants provided their results to an independent referee who collated and compared the data. The results are presented in an anonymised form together with discussion and conclusions which may be drawn from the exercise. Good agreement was obtained with standard deviations for individual measurements between 2.0% (54Mn) and 5.6% (60Co). Overall, the results obtained from the latter Round Robin show less consistency than those from the first Round Robin, which is attributed to the participation of a wider pool of organisations.

Following on from an initial Round Robin inter-comparison of gamma spectrometry measurements reported in 2014, this paper presents the results of the second part of a further Round Robin inter-comparison commissioned in 2018 by the European Working Group on Reactor Dosimetry. The purpose of the exercise was to demonstrate the level of consistency between different organisations’ measurements of the 93Nb(n,n’)93mNb reaction, which plays a key role in underwriting reactor dosimetry assessments. To achieve this, measurements of 93mNb activity were performed by twelve European organisations on six sets of near identical niobium samples, each having its own geometry and tantalum concentration. The samples were provided by CEA, France and irradiated in the MARIA reactor, operated by National Centre for Nuclear Research, Poland. Participants provided their results to an independent referee who collated and compared the data. The inter-comparison has demonstrated agreement to within standard deviations ranging from ±2.2% to ±7.9%, with a tendency for some organisations to measure elevated values. The results of the inter-comparison are presented in an anonymised form together with discussion and conclusions which may be drawn from the exercise.

In order to validate the neutron capture cross section in 241Am, comparisons of calculations with activation experiments at the TRIGA reactor at JSI and pile oscillation measurements at the MINERVE reactor at CEA Cadarache were done. The results suggest that the evaluation in the new test library JEFF-4T1 on average provide an improvement of the C/E results compared to the existing nuclear data library releases. However, due to some discrepancies, additional experiments at JSI as well as further development of the computational model of the JSI TRIGA reactor are planned to improve the accuracy and reliability of the neutron activation analysis and the nuclear data inferred from them.

The Cherenkov power meter developed at the Jožef Stefan Institute (JSI) is an independent, reliable and cost-effective measurement system based on Cherenkov radiation detection, used to measure fast power transients during pulse operation at the JSI TRIGA research reactor. It is based on a closed tube in order to avoid interference from external light sources and radiation damage experienced by optical fibers. The tube is placed in the reactor core periphery, the bottom part of the tube is filled with water, serving as the source of Cherenkov light. The measurements obtained in the framework of an extensive experimental campaign in collaboration with the French Atomic and Alternative Energy Commission (CEA) focused on reactor pulse operation using the Cherenkov power meter show excellent agreement with existing nuclear instrumentation (TRIGA pulse recorder), especially for high peak power pulses. However, the Cherenkov power meter outperforms the TRIGA pulse recorder in accurately recording low peak power pulses. The power vs. time behavior measured with the Cherenkov power meter is in accordance with measurements using miniature fission and ionization chambers developed at the CEA. The possibility of reactor pulse characterization based on measurements with the Cherenkov power meter combined with neutron dosimetry measurements sets the basis for enabling irradiation during pulse operation at JSI TRIGA research reactor in the future.

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.

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 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.

Constant improvements of the computational power and methods as well as demands of accurate and reliable measurements for reactor operation and safety require a continuous upgrade of the instrumentation. In particular, nuclear sensors used in nuclear fission reactors (research or power reactors) or in nuclear fusion facilities are operated under intense mixed neutron and gamma-ray fields, and need to be calibrated and modeled to provide selective and accurate neutron and gamma-ray measurements. The French Atomic Energy and Alternative Energies Commission (CEA) and the Jožef Stefan Institute (JSI) have started an experimental program dedicated to a detailed experimental benchmark with analysis using Monte Carlo particle transport calculations and a series of neutron and gamma-ray sensor types used in the JSI TRIGA Mark II reactor. CEA has setup a simplified TRIPOLI-4® modeling scheme of the JSI TRIGA reactor based on the information available in the IRPhEP benchmark in order to facilitate analysis of future neutron and gamma-ray measurements. These allow the CEA to perform a TRIPOLI-4 instrumentation calculation scheme benchmarked with the JSI MCNP model. This paper presents the main results of this CEA calculation scheme application and the analysis of their comparison to the JSI results obtained in 2012 with the MCNP5 & ENDF/B-VII.0 calculation scheme. This paper will conclude with some information about the new experimental program to be carried out in 2022 in the TRIGA reactor core.

In the 2019 C38 Deuterium-Deuterium campaign at JET several different ITER-relevant materials and dosimetry foils were irradiated in a specially designed long-term irradiation station located inside the vacuum vessel with the purpose of testing the activation of ITER materials by fusion neutrons. The samples were exposed to a neutron fluence of 1.9E14 n/cm2 during JET discharges performed in the experimental campaign over a period of 5 months. Gamma ray spectroscopy measurements were performed on irradiated samples to determine the activation of different long-lived isotopes in the samples. Monte Carlo computational analysis was performed to support the experiment by using the measured neutron yield and irradiation time. In this paper we focus on the computational analysis of the dosimetry foils that are used in order to measure the local neutron energy spectrum and flux. The foils were chosen to cover different neutron energies: thus Yttrium and some of the Nickel and Cobalt reactions were used to determine the Deuterium-Tritium fusion fraction, while Scandium and Iron and some of the Nickel and Cobalt reactions were used for comparison of the computed activity with the experimental measurements. The obtained C/E values show a reasonably good agreement between calculated and measured activity, thus validating the computational methodology and providing the basis for the analysis of the ITER-relevant materials and future experiments performed at JET in the Deuterium-Tritium campaign.

The availability of neutron fields with a high neutron flux, suitable for irradiation testing of nuclear instrumentation detectors relevant for applications in nuclear facilities such as material testing reactors (MTRs), nuclear power reactors and future fusion reactors is becoming increasingly limited. Over the last several years there has been increased interest in the experimental capabilities of the 250 kW Jožef Stefan Institute (JSI) TRIGA research reactor for such applications, however, the maximal achievable neutron flux in steady-state operation mode falls short of MTR-relevant conditions. The JSI TRIGA reactor can also operate in pulse mode, with a maximal achievable peak power of approximately 1 GW, for a duration of a few ms. A collaboration project between the JSI and the French Atomic and Alternative Energy Commission (CEA) was initiated to investigate absolute neutron flux measurements at very high neutron flux levels in reactor pulse operation. Such measurements will be made possible by special CEA-developed miniature fission chambers and modern data acquisition systems, supported by the JSI TRIGA instrumentation and activation dosimetry. Additionally, measurements of the intensity of Cherenkov light are proposed and being investigated as an alternative experimental method. This paper presents the preparatory activities for an exhaustive experimental campaign, which were carried out in 2019-2020, consisting of test measurements with not fully appropriate fission chambers, activation dosimetry and silicon photomultipliers (SiPMs) The presented results provide useful and promising experimental indications relevant for the design of the experimental campaign.

Neutron and gamma dose rate calculations were carried out around horizontal beam tube no. 5 at the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor. Results were compared to the experimental measurements in order to verify the computation model. In addition, another set of calculations and measurements was carried out, where an additional shield made out of concrete and paraffin was installed. With that configuration, we were able to study neutron and gamma scattering.

Activated cooling water in nuclear facilities can present a significant radiation source around primary cooling system causing radiation damage to electrical components, increasing doses to personnel and in the case of fusion facilities additional heating to superconducting coils. As there are only few sources of gamma rays with energies in the range of 6 MeV and 7 MeV an irradiation system using activated cooling water as the source of energetic gamma rays is proposed at the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor. Two different conceptual designs, one utilizing central irradiation channel and one utilizing radial piercing port for water activation, are presented and analysed in the paper. Despite an order of magnitude higher water activation in central channel compared to radial piercing port the 16N decay rate in the irradiation facility is comparable between both design (order of 108 decays per second) due to longer transient time from central channel to irradiation facility. In the irradiation facility the expected biological dose rates due to the 16N decay rate are in order of several mSv/h. From the results he conceptual design utilizing the radial piercing port currently presents the best option for the irradiation facility due to the simpler design of the irradiation loop, already present shielding of the loop and comparable number of 16N decay rates to central channel.

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), and by that way, to enhance border and port security barriers. Along the plan, material modification processes are employed firstly to study, and secondly to manipulate the most severe electrically active defects (which trap or annihilate free charge carriers), by specific ion implantation and defect engineering. This paper gives an overview of the experimental activities performed at the JSI TRIGA reactor in the framework of the E-SiCure project. Initial activities were aimed at obtaining information on the radiation hardness of SiC and at the study of the energy levels of the defects induced by neutron irradiation. Several Schottky barrier diodes were fabricated out of nitrogen-doped epitaxial grown 4H-SiC, and irradiated under Cd filters in the PT irradiation channel in the JSI TRIGA reactor with varying neutron fluence levels. Neutron-induced defects in the material were studied using temperature dependent current-voltage (I-V), capacitance-voltage (C-V) and Deep-Level Transient Spectroscopy (DLTS) measurements. Our prototype neutron detectors are configured as 4H-SiC-based Schottky barrier diodes for detection of secondary charged particles (tritons, alphas and lithium atoms) which are result of thermal neutron conversion process in 10B and 6LiF layers above the surface of the 4H-SiC diodes. For field testing of neutron detectors using a broad beam of reactor neutrons we designed a standalone prototype detection system consisting of a preamplifier, shaping amplifier and a multichannel analyser operated by a laptop computer. The reverse bias for the detector diode and the power to electronic system are 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.

Heating due to energy deposition of intense ionizing radiation in samples and structural materials of nuclear reactors poses severe limitations in terms of cooling requirements for safe reactor operation, especially in high neutron and gamma flux environments of material testing fission reactors (MTRs) and novel fusion devices. A bilateral CEA-JSI research project was launched in 2018 with the objective to measure the gamma heating rates in standard reactor-related materials (graphite, aluminium, stainless steel and tungsten) as well as fusionrelevant materials (low-activation steel Eurofer-97 and Nb3Sn superconductor) in the JSI TRIGA reactor my means of gamma calorimeters. The calorimeter design will be based on the the CALMOS-2 calorimeter developed at the CEA and used to perform gamma heating measurements in the OSIRIS MTR in Saclay. In order to optimize the detector response inside the JSI TRIGA reactor field and not to perturb the measurement field, a detailed computational analysis was performed in terms of energy deposition assessment and measurement field perturbation using the MCNP v6.1 code, and in terms of heat transfer using the COMSOL Multiphysics code. The abovementioned activities enabled us to finalize the detector design with the experimental campaign planned for the end of year 2019.

Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI) and Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA Cadarache). The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu) covered with thermal neutron filters (Cd and Gd). The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter) on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

Research reactors such as the “Jožzef Stefan” Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR’s structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.