Search

International audience; Fuel debris have a distinct neutron signature that can be detected to locate the said debris in a damaged nuclear power plant. Neutron measurement in a damaged PCV environment is however submitted to severe deployments constraints, including a high-dose-rate gamma background and limited available space. The study was therefore oriented towards small fission chambers (FC), with U-235-enriched active substrates. To investigate the expected performance of the FC in various irradiation conditions, a numerical model of the detector head was built. We describe the elaboration and experimental calibration of the numerical model and the Monte Carlo study of the fission rate inside U-235 coatings per generated neutron. The evaluation of a representative calibration coefficient then allowed us to carry out a multi-parameter performance study of a FC underwater, aiming at computing an explicit response function linking, on the one hand, the activity and spatial distribution of neutron emitters in a water container, with, one the other hand, the expected count rates measured by a fission chamber as a function of its radial and axial position inside the water volume. The FC underwater behavior was subsequently corroborated by a measurement campaign on a FC response, set at different positions inside a water drum, as a function of its axial and radial distance to a Cf-252 neutron source attached near the center of the container. We finally present an approach in which fuel debris localization is defined as an Inverse Problem, solvable with a Maximum-Likelihood Expectation Maximization (ML-EM) iterative algorithm. The projector matrix is built by capitalization on the results of the previously consolidated numerical studies. The ML-EM was tested on simulated data sets with a varying number of active voxels. Our first results indicate that, for a thermal neutron flux in the order of 10 n.cm−2.s−1 at the detector, originating voxels are identified with a spatial resolution in the radial plane in the order of 10 to 100 cm2.

International audience; This work aims at assessing the performance of a 235U enriched fission chamber in order to localize fuel debris, prior to dismantling operations, in a flooded primary containment vessel of a damaged nuclear reactor such as Fukushima Daiichi. Based on both a comprehensive scan of the environment and the detection of neutrons emitted by the melted core, fuel debris can be localized. In this paper, we carry out a simulation study using the MCNP6 code to investigate fission chamber response in the frame of fuel debris localization measurements in a damaged nuclear reactor. The CFUF34 fission chamber (manufactured by PHOTONIS) and the primary containment vessel of Fukushima Daiichi Unit 1 were chosen to conduct this work. Impact of different parameters were investigated with MCNP6, such as: neutron energy, water temperature, fission chamber position (altitude, lateral shift, and rotation), and sensitivity loss due to sediments potentially covering fuel debris. In summary, we show that fuel debris should be sought by their thermal neutron signature at a distance of a few centimeters and that potential rotational movements of the fission chamber up to 60° have a limited impact on signals measured. We also show that sensitivity loss due to sediments potentially covering fuel debris has been evaluated on the order of a factor 10 considering a 30 cm-thick sediment layer. On the other hand, experiments were performed to assess the impact of a strong gamma dose rate on fission chamber measurements. These irradiation trials involved a CFUE32 fission chamber (also manufactured by PHOTONIS) available in our laboratory and three different irradiation means: an X-ray tube, an 192Ir source, and a linear electron accelerator. These experiments enable to draw the conclusion that the fission chamber is not impacted by the gamma dose rate up to 104 Gy h−1, which is in good agreement with specifications provided by the manufacturer (PHOTONIS). In addition, no performance degradation was observed after an integrated gamma dose of 2200 Gy on the fission chamber in a 10 min irradiation. However, when the fission chamber is irradiated by gamma dose rates above 104 Gy h−1 (upper limit of the operating domain specified by PHOTONIS), a significant gamma background is observed. Nevertheless, as the gamma dose rates at Fukushima Daiichi should not exceed 103 Gy h−1, fission chamber measurements performed towards fuel debris localization in the primary containment vessels of the units would not be affected by the severe gamma-ray irradiation.

International audience; Monitoring fuel integrity during sodium-cooled fast reactor operation becomes a crucial issue when sodium uranoplutonate formed consecutively to a clad aperture will foster cooling defaults. Delayed neutron detection (DND) systems are the key tools allowing to prevent and to mitigate such an accident by monitoring the concentration of delayed neutron precursors contained into the coolant. DND systems operate in hard gamma-rays fluxes and low-neutron activity levels. They have to be carefully optimized in order to detect, as early as possible, potentially harmful failures. A discussion is conducted on privileged ways to design DND systems, and some conclusions are stated.

Natural gadolinium displays a number of remarkable physical properties: it is a rare earth element, composed of seven stable or quasi-stable isotopes, with an exceptionally high magnetization and a Curie point near room temperature. Its use in the field of nuclear instrumentation historically relates to its efficiency as a neutron poison in power reactors. Gadolinium is indeed the naturally occurring element with the highest interaction probability with neutrons at thermal energy, shared between Gd-157 (15.65%, 254000 b cross section) and Gd-155 (14.8%, 60900 b) isotopes. Considering that neutron capture results in an isotopic change, followed by a radiative rearrangement of nuclear and atomic structures, Gd may be embodied not merely as a neutron poison but as a neutron converter into a prompt photon and an electron source term. Depending on the nature and energy of the reaction products (from a few-keV Auger electrons up to 8 MeV gamma rays) that the detector aims at isolating as an indirect neutron signature, a variety of sensor media and counting methods have been introduced during the last decades. This review first draws a theoretical description of the radiative cascade following Gd ( n , γ ) capture. The cascade may be subdivided into regions of interest, each corresponding to dedicated detection designs and optimizations whose current status is detailed. This inventory has allowed the authors to extract and benchmark key figures of merit for the definition of a detection scheme: neutron attenuation, neutron sensitivity (cps/nv), gamma rejection, neutron detection limit in a mixed field, intrinsic or extrinsic moderation, and transportability. On this basis, the authors have identified promising paths for Gd-based neutron detection in contemporary instrumentation.

International audience; The use of a room-temperature gamma spectrometer is an issue in environmental radiation monitoring. To monitor radionuclides released around a nuclear power plant, suitable instruments giving fast and reliable information are required. High-pressure xenon (HPXe) chambers have range of resolution and efficiency equivalent to those of other medium resolution detectors such as those using NaI(Tl), CdZnTe, and LaBr3:Ce. An HPXe chamber could be a cost-effective alternative, assuming temperature stability and reliability. The CEA LIST actively studied and developed HPXe-based technology applied for environmental monitoring. Xenon purification and conditioning was performed. The design of a 4-L HPXe detector was performed to minimize the detector capacitance and the required power supply. Simulations were done with the MCNPX2.7 particle transport code to estimate the intrinsic efficiency of the HPXe detector. A behavioral study dealing with ballistic deficits and electronic noise will be utilized to provide perspective for further analysis.

International audience; To monitor radioactivity emitted by a vehicle such as a pedestrian, a car, a train, or a truck, radiation portal monitors (RPMs) are commonly employed. It has been noted that these RPM suffer from a shadow-shielding effect when the source is transported by a dense and large vehicle such as a truck or a train. While methods using databases are listed as state of the art, we have developed an approach based on a state model, ensuring a gain in performance and flexibility. This new study presents a technique applied to RPM, efficiently compensating the shadow-shielding effect. A simulation study has been performed to highlight the efficiency and reliability of the method.

Animma 2017 : Liège, Belgique, 19th to 23rd 2017; International audience; Delayed neutron detection systems are key apparati in Sodium cooled Fast Reactors to prevent safety accidents. Based on the feedbacks from Phenix and Superphenix, DND systems are studied by means of the Monte-Carlo particles transport code MCNP6. Promising designs are proposed and investigated as a result of the simulation study.

International audience; The development of portable and personal neutron dosimeters requires compact and efficient radiation sensors. Gd-157, Gd-155 and Cd-113 nuclei present the highest cross-sections for thermal neutron capture among natural isotopes. In order to allow for the exploitation of the low and medium-energy radiative signature of the said captures, the contribution of gamma background radiation, falling into the same energy range, needs to be cancelled out. This paper introduces a thermal neutron detector based on a twin-dense semiconductor scheme. The neutron-sensitive channel takes the form of a Gd-covered CdZnTe crystal, a high density and effective atomic number detection medium. The background compensation will be carried out by means of an identical CdZnTe sensor with a Tb cover. The setting of a hypothesis test aims at discriminating the signal generated by the signature of thermal neutron captures in Gd from statistical fluctuations over the compensation of both independent channels. The measurement campaign conducted with an integrated single-channel chain and two metal Gd and Tb covers, under Cs-137 and Cf-252 irradiations, provides first quantitative results on gamma rejection and neutron sensitivity. The described study of concept gives grounds for a portable, online compatible device, operable in conventional to controlled environments.

International audience; Nuclear instruments such as alpha/beta contamination meter are frequently used in a compensated mode where the contribution of gamma radiation background is compensated by a guard detector. The signal of interest is then the subtraction of counting from both channels. In practice, the noise signal measured by the guard detector is not strictly equal to the noise contribution into the first detector due to anisotropic biases.The random error (under Poisson assumption) is taken into account to build a hypothesis test. The system is also designed to minimize the systematic error but in some cases, this bias could not be completely removed. The measurement system then shows different behavior when the surrounding environment changes exhibiting inopportune false alarms.A method allowing the false alarms to be suppressed is addressed in this study for compensated measurement. An improvement in terms of reliability has been proven.

International audience; Online nuclear counting represents a challenge due to the stochastic nature of radioactivity. The counting data have to be filtered in order to provide a precise and accurate estimation of the count rate, while ensuring a response time compatible with the application in view. An innovative filter is presented in this paper to address this issue. The filter is nonlinear and based on a Centered Significance Test (CST) providing a local maximum likelihood estimation of the signal. This nonlinear approach allows enables to smooth the counting signal while maintaining a fast response when brutal change in activity occurs. The filter is then improved by the implementation of a Brown's double Exponential Smoothing (BES). The filter has been validated and compared to other state-of-the-art smoothing filters. The CST* filter shows a significant improvement compared to all tested smoothing filters.

International audience; Gadolinium-covered modified plastic scintillators show a high potential for the deployment of cost-effective neutron detectors. Taking advantage of the low-energy photon and electron signature of thermal neutron captures in gadolinium-155 and gadolinium-157 however requires a background correction. In order to display a trustable rate, dual compensation schemes appear as an alternative to Pulse Shape Discrimination. This paper presents the application of such a compensation scheme to a two-bismuth loaded plastic scintillator system. A detection scintillator interacts with incident photon and fast neutron radiations and is covered with a gadolinium converter to become thermal neutron-sensitive as well. In the meantime, an identical compensation scintillator, covered with terbium, solely interacts with the photon and fast neutron part of incident radiations. After the acquisition and the treatment of the counting signals from both sensors, a hypothesis test determines whether the resulting count rate after subtraction falls into statistical fluctuations or provides a robust image of neutron activity. A laboratory prototype is tested under both photon and neutron radiations, allowing us to investigate the performance of the overall compensation system. The study reveals satisfactory results in terms of robustness to a cesium-137 background and in terms of sensitivity in presence of a californium-252 source.

International audience; Plastic scintillator loading with gadolinium-rich organometallic complexes shows a high potential for the deployment of efficient and cost-effective neutron detectors. Due to the low-energy photon and electron signature of thermal neutron capture by Gd-155 and Gd-157, alternative treatment to pulse-shape discrimination has to be proposed in order to display a count rate. This paper discloses the principle of a compensation method applied to a two-scintillator system: a detection scintillator interacts with photon and fast neutron radiation and is loaded with gadolinium organometallic compound to become a thermal neutron absorber, while a not-gadolinium loaded compensation scintillator solely interacts with the fast neutron and photon part of incident radiation. After the nonlinear smoothing of the counting signals, a hypothesis test determines whether the resulting count rate post-background response compensation falls into statistical fluctuations or provides a robust indication of neutron activity. Laboratory samples are tested under both photon and neutron irradiations, allowing the authors to investigate the performance of the overall detection system in terms of sensitivity and detection limits, especially with regards to a similar-active volume He-3 based commercial counter. The study reveals satisfactory figures of merit in terms of sensitivity and directs future investigation toward promising paths.

Nuclear counting is a challenging task for nuclear instrumentation because of the stochastic nature of radioactivity. Event counting has to be processed and filtered to determine a stable count rate value and perform variation monitoring of the measured event. An innovative approach for nuclear counting is presented in this study, improving response time and maintaining count rate stability. Some nonlinear filters providing a local maximum likelihood estimation of the signal have been recently developed, which have been tested and compared with conventional linear filters. A nonlinear filter thus developed shows significant performance in terms of response time and measurement precision. The filter also presents the specificity of easy embedment into digital signal processor (DSP) electronics based on field-programmable gate arrays (FPGA) or microcontrollers, compatible with real-time requirements. © 2001 Elsevier Science. All rights reserved.

Correction for ‘Understanding the behaviour of different metals in loaded scintillators: discrepancy between gadolinium and bismuth’ by Guillaume H. V. Bertrand et al., J. Mater. Chem. C, 2015, 3, 6006.

Conference of 4th International Conference on Advancements in Nuclear Instrumentation Measurement Methods and their Applications, ANIMMA 2015 ; Conference Code:121554; International audience; Online nuclear counting represents a challenge due to the stochastic nature of radioactivity. The count data have to be filtered in order to provide a precise and accurate estimation of the count rate, this with a response time compatible with the application in view. An innovative filter is presented in this paper addressing this issue. It is a nonlinear filter based on a Centered Skellam Test (CST) giving a local maximum likelihood estimation of the signal based on a Poisson distribution assumption. This nonlinear approach allows to smooth the counting signal while maintaining a fast response when brutal change activity occur. The filter has been improved by the implementation of a Brown's double Exponential Smoothing (BES). The filter has been validated and compared to other state of the art smoothing filters. The CST-BES filter shows a significant improvement compared to all tested smoothing filters.

Conference of 4th International Conference on Advancements in Nuclear Instrumentation Measurement Methods and their Applications, ANIMMA 2015 ; Conference Code:121554; International audience; Plastic, scintillator loading with gadolinium-rich organometallic complexes shows a high potential for the deployment of efficient and cost-effective neutron detectors. Due to the low-energy photon and electron signature of thermal neutron capture by gadolinium-155 and gadolinium-157, alternative treatment to Pulse Shape Discrimination has to be proposed in order to display a trustable count rate. This paper discloses the principle of a compensation method applied to a two-scintillator system: a detection scintillator interacts with photon radiation and is loaded with gadolinium organometallic compound to become a thermal neutron absorber, while a non-gadolinium loaded compensation scintillator solely interacts with the photon part of the incident radiation. Posterior to the nonlinear smoothing of the counting signals, a hypothesis test determines whether the resulting count rate after photon response compensation falls into statistical fluctuations or provides a robust image of neutron activity. A laboratory prototype is tested under both photon and neutron irradiations, allowing us to investigate the performance of the overall compensation system in terms of neutron detection, especially with regards to a commercial helium-3 counter. The study reveals satisfactory results in terms of sensitivity and orientates future investigation toward promising axes.

Conference of 4th International Conference on Advancements in Nuclear Instrumentation Measurement Methods and their Applications, ANIMMA 2015 ; Conference Code:121554; International audience; Particle transport codes used in detector simulation allow the calculation of the energy deposited by charged particles produced following an interaction. The pulses temporal shaping is more and more used in nuclear measurement into pulse shape analysis techniques. A model is proposed in this paper to simulate the pulse temporal shaping and the associated noise level thanks to the output track file PTRAC provides by Monte-Carlo particle transport codes. The model has been dedicated to ion chambers and more especially for High Pressure Xenon chambers HPXe where the pulse shape analysis can resolve some issues regarding with this technology as the ballistic deficit phenomenon. The model is fully described and an example is presented as a validation of such full detector simulation.

International audience; Organometallic chemistry has recently gained a lot of attention in the domain of plastic scintillators. Homogenously dispersed metal complexes in a polymer matrix can afford plastic scintillators with unseen abilities. Heavy atom loading is very attractive as it gives access to plastics with increased sensitivity towards elusive radiations such as gamma and neutron. But this comes with a drawback, as heavy atoms tend to quench fluorescence, hence decreasing the scintillation yield. We present here a comprehensive study of this phenomenon with bismuth and gadolinium complexes. We investigate the influence of the ligand nature by varying organometallic and fluorophore concentration to probe their interaction. We also propose an explanation of the difference in behavior between these two metals. These results were applied to the fabrication of large volume loaded plastic scintillators (>100 cm3). Bismuth loaded scintillators displayed characteristics equivalent to lead loaded commercial materials, and gadolinium samples proved to be able to capture thermal neutrons and release gamma rays.

International audience; This article presents the synthesis and the blend of bismuth complexes in polystyrene based plastic scintillators. A specific design has enabled the fabrication of a scintillator loaded with up to 17 wt% of bismuth. Tri-carboxylate and triaryl bismuth compounds were used to explore and understand the influence of bismuth loading on the two main criteria of plastic scintillation: light yield and detection efficiency of gamma-rays. For gamma radiation with an energy

Conference of 2013 3rd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and Their Applications, ANIMMA 2013 ; Conference Code:102802; International audience; To monitor radioactivity passing through a pipe or in a given container such as a train or a truck, radiation detection systems are commonly employed. These detectors could be used in a network set along the source track to increase the overall detection efficiency. However detection methods are based on counting statistics analysis. The method usually implemented consists in trigging an alarm when an individual signal rises over a threshold initially estimated in regards to the natural background signal. The detection efficiency is then proportional to the number of detectors in use, due to the fact that each sensor is taken as a standalone sensor. A new approach is presented in this paper taking into account the temporal periodicity of the signals taken by all distributed sensors as a whole. This detection method is not based only on counting statistics but also on the temporal series analysis aspect. Therefore, a specific algorithm is then developed in our lab for this kind of applications and shows a significant improvement, especially in terms of detection efficiency and false alarms reduction. We also plan on extracting information from the source vector. This paper presents the theoretical approach and some preliminary results obtain in our laboratory.

Illegal radioactive material transportation detection, by terrorist for example, is problematic in urban public transportation. Academics and industrials systems include Radiation Portal Monitor (RPM) to detect radioactive matters transported in vehicles or carried by pedestrians. However, today's RPMs are not able to efficiently detect a radioactive material in movement. Due to count statistic and gamma background, false alarms may be triggered or at the contrary a radioactive material not detected. The statistical false alarm rate has to be as low as possible in order to limit useless intervention especially in urban mass transportation. The real-time approach depicted in this paper consists in using a time correlated detection technique in association with a sensor network. It is based on several low-cost and large area plastic scintillators and a digital signal processing designed for signal reconstruction from the sensor network. The number of sensors used in the network can be adapted to fit with applications requirements or cost. The reconstructed signal is improved by comparing other approaches. This allows us to increase the device speed that has to be scanned while decreasing the risk of false alarm. In the framework of a project called SECUR-ED Secured Urban Transportation - European Demonstration, this prototype system will be used during an experiment in the Milan urban mass transportation.

Conference of 2013 3rd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and Their Applications, ANIMMA 2013 ; Conference Code:102802; International audience; Counting measurements associated with nuclear instruments are tricky to carry out due to the stochastic process of the radioactivity. Indeed events counting have to be processed and filtered in order to display a stable count rate value and to allow variations monitoring in the measured activity. Smoothers (as the moving average) are adjusted by a time constant defined as a compromise between stability and response time. A new approach has been developed and consists in improving the response time while maintaining count rate stability. It uses the combination of a smoother together with a detection filter. A memory of counting data is processed to calculate several count rate estimates using several integration times. These estimates are then sorted into the memory from short to long integration times. A measurement position, in terms of integration time, is then chosen into this memory after a detection test. An inhomogeneity into the Poisson counting process is detected by comparison between current position estimate and the other estimates contained into the memory in respect with the associated statistical variance calculated with homogeneous assumption. The measurement position (historical time) and the ability to forget an obsolete data or to keep in memory a useful data are managed using the detection test result. The proposed smoother is then an adaptive and a learning algorithm allowing an optimization of the response time while maintaining measurement counting stability and converging efficiently to the best counting estimate after an effective change in activity. This algorithm has also the specificity to be low recursive and thus easily embedded into DSP electronics based on FPGA or micro-controllers meeting "real life" time requirements.

Conference of 2013 3rd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and Their Applications, ANIMMA 2013 ; Conference Code:102802; International audience; War against CBRN-E threats needs to continuously develop sensors with improved detection efficiency. More particularly, this topic concerns the NR controls for homeland security. A first analysis requires indeed a fast gamma spectrometry so as to detect potential special nuclear materials (SNM). To this aim, plastic scintillators could represent the best alternative for the production of large-scale, low-cost radiation portal monitors to be deployed on boarders, tolls, etc. Although they are known to be highly sensitive to gamma rays, due to their poor resolution, information relative to the nature of the SNM is tricky. Thus, only the Compton edge is obtained after interaction, and no information of the photoelectric peak is observed. This project concerns new developments on a possible pseudo-gamma spectrometry performed with plastic scintillators. This project is articulated on a combination of two developments: The design of new materials most suitable for recovering the photoelectric peak after gamma interaction with the scintillator. This work concerns mainly plastic scintillators loading with heavy elements, such as lead or bismuth. The analysis of the resulting signal with smart algorithms. This work is thus a pluridisciplinary work performed at CEA LIST and embeds 4 main disciplines: MCNPX simulations (simulated spectra), chemistry of materials (preparation of various plastic scintillators with different properties), instrumentation (lab experiments) and smart algorithms. Really impressive results were obtained with the unfolding of simulated spectra at various energies (from 241Am to 60Co) and an innovative approach was proposed to counter-balance the quenching effect of luminescence by heavy elements in plastic scintillators.

Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,α) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phénix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

Fourth generation reactors are being developed for renewing the nuclear energy industry with safer reactors that optimize uranium resources and produce less long-lived radioactive waste. France has focused on the development of SFR “Sodium Fast Reactor”. In the past, prototypes have been built and operated such as Rapsodie, Phénix, Superphénix. Since 2006, the CEA (French Atomic Energy Commission) has overseen the design of the future prototype of this technology: ASTRID. It is within this framework that teams at the French Atomic Energy Commission are working on the development of instrumental means to guarantee the absolute integrity of the first containment barrier: the fuel cladding. Indeed, the safety level to be achieved in the fourth-generation program requires the operator to continuously monitor the fission products released into the primary coolant. To ensure this safety function, various systems were implemented, such as delayed neutron detection (DND) and gamma spectrometry systems. These systems have been upgraded since their first operations through R&D programs in line with technological advances and new digital design methods.The historical gamma spectrometry instrumentation for clad failure monitoring is compromised by the presence of significant background that leads to delay lines of about 15 minutes necessary for the 23Ne (T1/2 =38 s) decay. In addition, the other background source, i.e. 41Ar, with a half-life of 110 minutes, was not filtered by this instrumentation. Indeed, it loses only 7 % of its activity in these delay lines.Regarding the safety requirement set for SFR, we will study in the frame of this work the potential contribution of a low background instrumentation by means of a Compton suppression system integrated into the SFR measuring device. The main goal of such an instrumentation would be, first, to strengthen the diagnostic of the detection of short-period fission products (about 3 min) which could indicate an early contact between the fuel and the coolant that can be dangerous for the reactor safety, second, to ensure a fast detection by improving the measuring response time.First, the source term of fission products was characterized by putting into equation various physical phenomena that govern the behavior of fission gas in an SFR environment following a clad failure. Expected fission products activities have been estimated. Second, we performed a numerical study, by means of the Monte Carlo method, of a low background instrumentation, from its experimental validation to its optimization. Finally, we implemented the model for the problem of cladding failure in the case of SFR environment. Minimum detectable activities have been estimated then compared to expected activities. The added value of this instrumentation for applications to the early detection of clad failures in an SFR environment will then be exposed.; Les réacteurs de quatrième génération sont développés dans le but de renouveler la filière énergétique du nucléaire avec des réacteurs plus sûrs, optimisant la consommation du minerai uranium et produisant moins de déchets de vie longue. La France a investi dans le développement de la technologie des réacteurs à neutrons rapides refroidis au sodium. Par le passé, des prototypes ont été construits et exploités tels que RAPSODIE, PHENIX, SUPERPHENIX ; et depuis 2006, le CEA est en charge de la conception du futur prototype de cette filière : ASTRID. C’est dans ce cadre que les équipes du Commissariat à l’énergie atomique travaillent sur l’élaboration de moyens instrumentaux permettant de garantir l’intégrité absolue de la première barrière de confinement : la gaine des aiguilles combustibles. En effet, le niveau de sûreté à atteindre dans le programme génération 4 oblige l’exploitant à surveiller en continu la propreté du fluide caloporteur primaire. Pour assurer cette fonction de sûreté, différents systèmes sont implémentés tels que les systèmes de détection des neutrons différés (DND) et les systèmes de spectrométrie gamma. Ces systèmes sont en continuelle mutation depuis leurs premières exploitations par des programmes R&D assurant la prise en compte du retour d’exploitation de ces instruments et la remise à niveau en regard des avancées technologiques et des nouveaux moyens numériques de conception. L’instrumentation historique pour le suivi isotopique des gaz de fission par spectrométrie gamma est très bruitée et est compromise par la présence de lignes à retard d’environ 15 minutes nécessaires à la désactivation du 23Ne (T1/2 =38 s). De plus, l’autre source de bruit, i.e. l’41Ar, présentant une période de 110 minutes, n’a pas été filtrée par cette instrumentation. En effet, ce dernier ne perd que 7% de son activité dans ces lignes à retard.Au regard des exigences de sûreté fixées pour les RNR-Na, nous étudierons dans le cadre de cette thèse l’apport potentiel d’une instrumentation bas bruit au moyen d’un système à suppression Compton intégré au poste de mesure RNR-Na. L’enjeu majeur d’une telle instrumentation serait, d’une part, d’enrichir le diagnostic par la détection des produits de fission de courte période (~3 min), lesquels peuvent signer précocement une mise en contact entre le combustible et le caloporteur dangereuse pour la sûreté du réacteur, d’autre part, d’assurer une détection rapide par l’amélioration du temps de réponse de la mesure.Dans un premier temps, le terme source des produits de fission a été caractérisé par la mise en équation des différents phénomènes physiques régissant le comportement des gaz de fission dans un environnement de type RNR-Na faisant suite à une rupture de gaine. Des activités attendues des produits de fission sont ainsi estimées. Dans un second temps, nous avons procédé à l’étude numérique, au moyen de la méthode Monte Carlo, d’une instrumentation bas bruit, à sa validation expérimentale, puis à son optimisation. Enfin, nous avons implémenté la modélisation de la mesure spectrométrique bas bruit pour la problématique des ruptures de gaine dans le cadre d’un environnement RNR-Na. Des activités minimales détectables de produits de fission sont, d’abord, calculées puis comparées aux activités attendues. La plus-value de cette instrumentation pour l’application à la détection précoce des pertes d’étanchéité dans un environnement type RNR-Na sera alors exposée.

Fourth generation reactors are being developed for renewing the nuclear energy industry with safer reactors that optimize uranium resources and produce less long-lived radioactive waste. France has focused on the development of SFR “Sodium Fast Reactor”. In the past, prototypes have been built and operated such as Rapsodie, Phénix, Superphénix. Since 2006, the CEA (French Atomic Energy Commission) has overseen the design of the future prototype of this technology: ASTRID. It is within this framework that teams at the French Atomic Energy Commission are working on the development of instrumental means to guarantee the absolute integrity of the first containment barrier: the fuel cladding. Indeed, the safety level to be achieved in the fourth-generation program requires the operator to continuously monitor the fission products released into the primary coolant. To ensure this safety function, various systems were implemented, such as delayed neutron detection (DND) and gamma spectrometry systems. These systems have been upgraded since their first operations through R&D programs in line with technological advances and new digital design methods.The historical gamma spectrometry instrumentation for clad failure monitoring is compromised by the presence of significant background that leads to delay lines of about 15 minutes necessary for the 23Ne (T1/2 =38 s) decay. In addition, the other background source, i.e. 41Ar, with a half-life of 110 minutes, was not filtered by this instrumentation. Indeed, it loses only 7 % of its activity in these delay lines.Regarding the safety requirement set for SFR, we will study in the frame of this work the potential contribution of a low background instrumentation by means of a Compton suppression system integrated into the SFR measuring device. The main goal of such an instrumentation would be, first, to strengthen the diagnostic of the detection of short-period fission products (about 3 min) which could indicate an early contact between the fuel and the coolant that can be dangerous for the reactor safety, second, to ensure a fast detection by improving the measuring response time.First, the source term of fission products was characterized by putting into equation various physical phenomena that govern the behavior of fission gas in an SFR environment following a clad failure. Expected fission products activities have been estimated. Second, we performed a numerical study, by means of the Monte Carlo method, of a low background instrumentation, from its experimental validation to its optimization. Finally, we implemented the model for the problem of cladding failure in the case of SFR environment. Minimum detectable activities have been estimated then compared to expected activities. The added value of this instrumentation for applications to the early detection of clad failures in an SFR environment will then be exposed.; Les réacteurs de quatrième génération sont développés dans le but de renouveler la filière énergétique du nucléaire avec des réacteurs plus sûrs, optimisant la consommation du minerai uranium et produisant moins de déchets de vie longue. La France a investi dans le développement de la technologie des réacteurs à neutrons rapides refroidis au sodium. Par le passé, des prototypes ont été construits et exploités tels que RAPSODIE, PHENIX, SUPERPHENIX ; et depuis 2006, le CEA est en charge de la conception du futur prototype de cette filière : ASTRID. C’est dans ce cadre que les équipes du Commissariat à l’énergie atomique travaillent sur l’élaboration de moyens instrumentaux permettant de garantir l’intégrité absolue de la première barrière de confinement : la gaine des aiguilles combustibles. En effet, le niveau de sûreté à atteindre dans le programme génération 4 oblige l’exploitant à surveiller en continu la propreté du fluide caloporteur primaire. Pour assurer cette fonction de sûreté, différents systèmes sont implémentés tels que les systèmes de détection des neutrons différés (DND) et les systèmes de spectrométrie gamma. Ces systèmes sont en continuelle mutation depuis leurs premières exploitations par des programmes R&D assurant la prise en compte du retour d’exploitation de ces instruments et la remise à niveau en regard des avancées technologiques et des nouveaux moyens numériques de conception. L’instrumentation historique pour le suivi isotopique des gaz de fission par spectrométrie gamma est très bruitée et est compromise par la présence de lignes à retard d’environ 15 minutes nécessaires à la désactivation du 23Ne (T1/2 =38 s). De plus, l’autre source de bruit, i.e. l’41Ar, présentant une période de 110 minutes, n’a pas été filtrée par cette instrumentation. En effet, ce dernier ne perd que 7% de son activité dans ces lignes à retard.Au regard des exigences de sûreté fixées pour les RNR-Na, nous étudierons dans le cadre de cette thèse l’apport potentiel d’une instrumentation bas bruit au moyen d’un système à suppression Compton intégré au poste de mesure RNR-Na. L’enjeu majeur d’une telle instrumentation serait, d’une part, d’enrichir le diagnostic par la détection des produits de fission de courte période (~3 min), lesquels peuvent signer précocement une mise en contact entre le combustible et le caloporteur dangereuse pour la sûreté du réacteur, d’autre part, d’assurer une détection rapide par l’amélioration du temps de réponse de la mesure.Dans un premier temps, le terme source des produits de fission a été caractérisé par la mise en équation des différents phénomènes physiques régissant le comportement des gaz de fission dans un environnement de type RNR-Na faisant suite à une rupture de gaine. Des activités attendues des produits de fission sont ainsi estimées. Dans un second temps, nous avons procédé à l’étude numérique, au moyen de la méthode Monte Carlo, d’une instrumentation bas bruit, à sa validation expérimentale, puis à son optimisation. Enfin, nous avons implémenté la modélisation de la mesure spectrométrique bas bruit pour la problématique des ruptures de gaine dans le cadre d’un environnement RNR-Na. Des activités minimales détectables de produits de fission sont, d’abord, calculées puis comparées aux activités attendues. La plus-value de cette instrumentation pour l’application à la détection précoce des pertes d’étanchéité dans un environnement type RNR-Na sera alors exposée.