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The SSH Open Marketplace is a discovery portal for digital tools, services, and methods in the Social Sciences and Humanities (SSH), with an emphasis placed on placing these resources in their research context. It is the SSH component of the EOSC Marketplace. Previous initiatives have shown that curation of content is the key element to maintain the attractiveness and the utility of a discovery portal like the SSH Open Marketplace.1 Thus at each stage of the development of the Marketplace, from the first ingestion of existing catalogues to the curation sprint, curation has been at the heart of conception. This emphasis, based on previous examples and the work done throughout the SSHOC project, has led to profound changes in the data model, definition of quality criteria, standardization of the content as the team adapts the machinery of the Marketplace based on lessons learned. This continuous work has laid the foundation of the development of Editorial Guidelines which will be used by future curation actors as well as the definition of their different curation roles and the organisation of their relationship.

article soumis à 26TH INTERNATIONAL CONFERENCE ON OPTICAL FIBRE SENSORS OFS-26 et refusé par la conférence; International audience; The DISCOMS project and its issues, regarding current scientific challenges and potential benefits for nuclear plant safety in case of severe accident with reactor vessel failure, have been discussed. We claim such remote monitoring solution based on state-of-the-art of quasi-distributed OFS and specific long-length SPND should greatly contribute to enhance the quality of real time information during such event, providing first responders and nuclear safety people with direct data concerning both the corium location and the health of the 3rd barrier of confinement. In such context, the monitoring system developed in the scope of the DISCOMS project will enable the crisis teams in charge of handling the emergency situations on NPP to reduce the uncertainties associated to some tools currently used for severe accident mitigation (e.g. computations aids based on calculations of pre-defined Severe Accident scenarios). In particular, after the reactor pressure vessel failure, DISCOMS monitoring will provide the precise and real-time view of the progression front of core melt inside the reactor building basemat or the core catcher (if any). Based on this actual representation of the basemat ablation, the crisis team will be able to optimize the recommended mitigation actions, both for accident management on-site and population protection. In April 18, 2016 the International Atomic Energy Agency records 444 Nuclear Power Plants (NPPs) currently in operation and 64 under construction, located in 31 countries, which means the potential market for the DISCOMS solution is worldwide.

National audience; Ce papier presente trois etapes relatives a l'etude d'un nouveau calorimetre differentiel concu pour la mesure du debit de dose adsorbee dans des canaux experimentaux en reacteur nucleaire d'irradiations la conception du design des cellules calorimetriques par simulation numerique en thermique, l'etalonnage preliminaire hors reacteur des cellules calorimetriques plus compactes en montage differentiel et la qualification de ce calorimetre lors d'une campagne d'irradiations dans le reacteur MARIA (comparaison avec un calorimetre differentiel classique

International audience

International audience

International audience

International audience

International audience; A numerical toolbox based on Monte Carlo calculations for self-powered neutron and gamma detector design, simulation and operation has been developed in order to save operators numerous and fastidious calibration tests. The experimental validation of this numerical toolbox, called MATiSSe', in the OSIRIS Material Testing Reactor is fully detailed in the paper. C/E comparisons presuppose promising future application for the numerical toolbox.

International audience; The benchmark experiments performed in an operating research reactor cannot achieve the same level of accuracy as benchmarks in dedicated facilities that are specifically designed for such a purpose. However research reactors offer a great opportunity for benchmark experiments when designed and performed with great caution and accuracy. The paper describes a series of experiments performed at the JSI TRIGA reactor that can serve as benchmark experiments for testing computer codes and nuclear data. The experiments described are criticality, Au (n,g) and Al (n,a) reaction rates in irradiation channels, absolute and relative Au (n,g), $^{235}$U (n,f) $^{238}$U(n,f) reaction rates in the core, burnup, control rod worth, isothermal reactivity coefficient, self-shielding. Some of the experiments have already been evaluated and are available to worldwide community, while the others are yet to be evaluated.

International audience; In order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its RandD effort in the field of in-pile instrumentation during the last decade. Major progresses have thus been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. Benefits of this enhanced measurement potential, illustrated with some improvements achieved in neutron and gamma flux detection, are described in this paper.This paper also discusses the needs for irradiation capabilities within the coming years, in order to complete the in-pile qualification of newly developed sensors and the extensive validation of their corresponding simulation model.

International audience; At the Reactor Physics Department of the Jozef Stefan Institute (JSI) a detailed Monte Carlogeometrical model of the TRIGA Mark II research reactor has been developed. It has beencontinuously utilized for the calculation of reactor parameters and in the support of the designand safety assessment of various experiments. A series of experiments aimed at measuringthe core neutron flux and reaction rate distributions has been performed with the objective of comparing the measurements and Monte Carlo calculations - the ultimate goal being the validation of the computational model.

International audience; The Fukushima-Daiichi nuclear disaster showed that the need for safety must always prevail. This paper discusses the development of remote monitoring technologies to improve Nuclear Power Plants (NPPs) safety, in operation (Pressurized Water Reactors), under construction (the EPR reactors, i.e. the GEN 3 PWR), or for any other next generations of reactors. At Fukushima, the total loss of electrical power supplies has quickly led most of the instrumentation inoperative and the operator (TEPCO) with no way to monitor the status and the evolution of the accident. To overcome these important drawbacks, advantage can be taken from the considerable potential of distributed sensing technologies based on both "Optical Fiber Sensors" (Raman, Brillouin, and Rayleigh Reflectometries) and long-length "Self Powered Neutron Detectors" (SPNDs). The goal consists in inquiring about the status of the third barrier of confinement and to define possible mitigation strategies in case of severe accident, namely: i) reactor pressure vessel breakthrough and corium relocation outside the vessel, ii) concrete basemat erosion and iii) corium cooling. Such monitoring should consist in "sensing cables" embedded in concrete basemat below the reactor vessel and interrogated from a rear base where operators can work safely. In this context, DISCOMS, which stands for "DIstributed Sensing for COrium Monitoring and Safety", is a five-year project, managed by the French National Research Agency (ANR), dealing with the NPP safety improvement, from normal situation to severe accidents. Monitoring phases include reactor vessel breaching, corium flow, along with post-accidental period (corium cooling ex-vessel). Thus, optical fibers selected for their resistance to ionizing radiations and long length SPNDs, both judiciously deployed within the reactor concrete basemat, and the structures around it, will provide a useful real-time or on-demand monitoring, in normal operation, and more important in accidental and post-accidental situations.

International audience; The Fukushima-Daiichi nuclear disaster showed that the need for safety must always prevail. This paper discusses the development of remote monitoring technologies to improve Nuclear Power Plants (NPPs) safety, in operation (Pressurized Water Reactors), under construction (the EPR reactors, i.e. the GEN 3 PWR), or for any other next generations of reactors. At Fukushima, the total loss of electrical power supplies has quickly led most of the instrumentation inoperative and the operator (TEPCO) with no way to monitor the status and the evolution of the accident. To overcome these important drawbacks, advantage can be taken from the considerable potential of distributed sensing technologies based on both "Optical Fiber Sensors" (Raman, Brillouin, and Rayleigh Reflectometries) and long-length "Self Powered Neutron Detectors" (SPNDs). The goal consists in inquiring about the status of the third barrier of confinement and to define possible mitigation strategies in case of severe accident, namely: i) reactor pressure vessel breakthrough and corium relocation outside the vessel, ii) concrete basemat erosion and iii) corium cooling. Such monitoring should consist in "sensing cables" embedded in concrete basemat below the reactor vessel and interrogated from a rear base where operators can work safely. In this context, DISCOMS, which stands for "DIstributed Sensing for COrium Monitoring and Safety", is a five-year project, managed by the French National Research Agency (ANR), dealing with the NPP safety improvement, from normal situation to severe accidents. Monitoring phases include reactor vessel breaching, corium flow, along with post-accidental period (corium cooling ex-vessel). Thus, optical fibers selected for their resistance to ionizing radiations and long length SPNDs, both judiciously deployed within the reactor concrete basemat, and the structures around it, will provide a useful real-time or on-demand monitoring, in normal operation, and more important in accidental and post-accidental situations.

Cet article est issu de la conférence "ANIMMA 2017" (Advancements in Nuclear Instrumentation Measurements Methods and their Applications) tenue à Liège du 19 au 23 juin 2017; International audience; In order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its research and development effort in the field of in-pile instrumentation during the last decade. Major progresses have thus been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. A significant part of this effort have been conducted in the framework of the Joint Instrumentation Laboratory between the CEA and the Belgian Nuclear Research Centre (SCKCEN).In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile. The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 10$^{15}$ n.cm$^{-2}$.s$^{-1}$ and is generally dominated by thermal neutrons. Moreover, the neutron flux is accompanied by a high gamma flux of typically a few 10$^{15}$$\gamma$cm$^{-2}$.s$^{-1}$, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors allowing the simultaneous detection of both thermal and fast neutron flux. Thermal neutrons can be measured using a Self-Powered-Neutron-Detector (SPND) or a $^{235}$U miniature fission chamber, while fast neutron detection requires a miniature fission chamber with a special fissile material presenting an appropriate energy threshold, which can be $^{242}$Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCKCEN in Belgium. Two FNDS prototypes were operated in-pile during nearly 1000 hours. These tests exhibited the consistency of the measurement of thermal to fast neutron flux ratio with MCNP calculations, as well as the right compensation of fissile material depletion. Then a second test was recently completed at ISIS reactor operated by CEA in France. For this irradiation, FNDS signal was compared to reference thermal and fast neutron flux measurements using activation dosimeters analyzed under COFRAC Quality Certification. During this latter test, FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.This paper describes the innovative features of FNDS and discusses the results of its final in-pile qualification. FNDS is now operational and is assumed to be the first and unique acquisition system able to provide an online measurement of the fast neutron flux in MTR conditions. This system will of course be used to perform spectral neutron characterization of JHR channels, but it may also be implemented in future irradiation experiments, for a better and real-time evaluation of the fast neutron flux received by material and fuel samples.

International audience; The paper focuses on a validation experiment performed at the JSI TRIGA research reactor using two absolutely calibrated CEA developed miniature fission chambers. These were used to measure a set of axial fission rate profiles through the height of the reactor core which were compared to reaction rate profiles computed using the Monte Carlo method. The purpose of the paper is to describe the support calculations and methodology used in the evaluation of the experimental and computational biases and uncertainties. This includes the study of the effect of the fission chamber structure and the aluminum guide tubes used for in-core insertion on the fission rate measurements as well as a detailed description of the influence of fuel material data error and fuel burn-up on the experimental uncertainty and computational biases. It is shown that the insertion of fission chambers and guide tubes into the core of the TRIGA reactor significantly influences the local neutron spectrum and consequently the fission rate measurements one can observe an up to 15 % decrease. A similar effect was recorded when modelling the fuel burnup, where a 10 % deficit of $^{235}$U mass results in an approximately 10 % decrease in fission rates

International audience; The Phénix nuclear power plant has been a French Sodium Fast Reactor (SFR) prototype producing electrical power between 1973 and 2010. The power was monitored using ex-core neutron measurements. This kind of measurement instantly estimates the power but needs to be often calibrated with the heat balance thermodynamic measurement. Large safety and security margins have then been set not to derive above the nominal operating point. It is important for future SFR to reduce this margin and working closer to the nominal operating point. This work deals with the use of delayed gamma to measure the power. The main activation product contained in the primary sodium coolant is the 24Na which is not convenient for neutron flux measurement due to its long decay period. The experimental study done at the Phénix reactor shows that the use of 20F as power tagging agent gives a fast and accurate power measurement closed to the thermal balance measurement thanks to its high energy photon emission (1.634 MeV) and its short decay period (11 s).

On sodium fast reactors, power is usually measured by heat balance on water to vapor heat exchangers which is correlated with the ex-core neutron flux measurements. The ex-core neutron power measurement is done by fission chambers (five on the French SFR Phnix reactor) with several activity ranges settled at the bottom of the primary vessel to cover the whole neutron flux range of more than 11 decades. This measurement instantly estimates the neutron power but induces some shift problems. The clad failure detection is done by gamma measurement on argon cover gas sampling and neutron measurement on primary sodium sample. This work deals with the use of gamma spectrometry for fourth generation SFR power monitoring and clad failure detection. Usually gamma spectrometers could not manage on-line application but recent improvements in this research field may improve it. The Adonis digital signal processing prototype gives a time function response and is able to do spectrometry measurement with metrological grade under high gamma flux impinging the HPGe detector, with also high and fast activity transient. Previous works on Pressurized Water Reactors (PWR) show that gamma emitter concentration in primary coolant is also directly correlated with reactor neutron power. On SFR, the use of short decay period gamma emitters as the 20F and 23Ne tagging agent will allow a very fast response system without thermal hydraulic distortion effects. The study is divided in two parts. First part deals with a simulation study based on physical calculations and MCNP simulations to predict sensor signal as a function of neutron power. The second part is an experimental test under preparation at the French SFR Phnix. Measuring activation and fission products contained in primary sodium, the adaptive Adonis system set on primary coolant sample could be an innovating neutron power monitoring and clad failure detection system for SFR.

International audience; This work deals with the use of gamma spectrometry to monitor the fourth generation of sodium fast reactor (SFR) power. Simulation part has shown that power monitoring in short response time and with good accuracy is possible measuring liquid sodium delayed gamma emitters produced in-core. An experimental test is under preparation at French SFR Phénix experimental reactor to validate simulation studies. Physical calculations have been done to correlate gamma activity to the released thermal power. Gamma emitter production rate in the reactor core was calculated with technical and nuclear data as sodium velocity, atomic densities, neutron spectra and incident neutron cross-sections of reactions producing gamma emitters. Then, a thermal hydraulic transfer function was used for taking into account primary sodium flow in our calculations and gamma spectra were determined by Monte-Carlo simulations. For power monitoring problematic, use of a short decay period gamma emitter will allowed to obtain a very fast response system without cumulative and flow distortion effects. The experiment will be set during the reactor "end of life testing". The Delayed Neutron Detection (DND) system cell has been chosen as the best available primary sodium sample for gamma power monitoring on Phénix reactor due to its short transit time from reactor core to measurement sample and homogenized sampling in the reactor hot pool. The main gamma spectrometer is composed of a coaxial high purity germanium diode (HPGe) coupled with a transistor reset preamplifier. The signal will be then processed by a digital signal processing system (called Adonis) which is optimum for high count rate and various time activity measurements. To limit statistical problems of the signal, an analytical pileup correction method using duration variable given by our spectrometry system Adonis) and a nonparametric Bayesian inference for photopeack deconvolution will be used.