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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 investigates a compact Compton Suppression System (CSS) based on a High-Purity Germanium (HPGe) diode as a primary spectrometer. From a wide range of potential materials for the guard detector, including plastic scintillator, Thallium doped Sodium Iodide (NaI(Tl)), and Lanthanum III bromide (LaBr3), a bismuth germanate crystal (BGO) is selected on account of a high photon mass-attenuation coefficient, thus allowing more compactness. The authors then describe the complete design study of an HPGe/BGO CSS based on the post-processing of Monte Carlo particle track history files (MCNP6.1/PTRAC). The calculation program, simulating the filtering of coincident events on both primary and secondary detectors, is validated through a series of experimental measurements using Am-241, Cs-137, and Co-60 sources in various configurations. On the basis of this empirically validated calculation, an optimization study of the BGO-active shield is performed over two factors of merit: Compton Suppression Factor (CSF) and Full-Energy Peak Efficiency. Optimization steps subsequently involve the scaling of BGO internal and external radii, the thickness of the entrance window, and the diameter of a collimation hole. Eventually, performance figures of the CSS are illustrated in a case study involving the characterization of large waste containers, carrying low levels of uranium activity to seek among natural radiological background. Calculation results show a lowering of minimal detectable activities (MDA) by a factor of 2.5 based on the signature of 1001-keV gamma rays from U-238.

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.

The principle of the proposed method is based on a simple and efficient way to estimate the muon energy deposition in a cylindrical sensor. Muon initial energy is randomly sampled from experimental distributions and the traveled distance through the scintillator volume is assessed from a Monte-Carlo model in cylindrical coordinates. Energy deposition is estimated by solving Bethe-Block formula thanks to digital Euler approximation. This model has been experimentally tested in our laboratory using NaI(Tl) scintillator apparatus. Results from this comparison show good agreement and confirm the potential of the model. Sources of discrepancy are emphasized, while explaining peak shift between experimental results and theoretical calculations.

Pulse shape analysis is commonly implemented into nuclear detection systems notably dealing with phoswich structures addressingfor instance beta or neutron discrimination from gammarays. Particle transport codes such as MCNP, TRIPOLI or GEANT are useful toolsto help designing phoswich systems. However, the basic calculation is focused on the estimation of the deposited energy per particlehistory into scintillator parts. The new model proposed by the CEA LIST enables to process scintillation light production, collection, conversion, and thereafter pulse shape simulation at the output of a photomultiplier.

The approach of the scheme is to pile-up X-ray photons on the scintillator to generate a high energy peak in the spectrum. The system includes a miniature X-ray generator, a brass disk with a thin slit rotated by a motor and placed in front of the X-ray window, and a collimator placed in front of the scintillation detector to reduce the time of alignment between the scintillation detector and the X-ray generator. The rotating brass disk placed between the X-ray source and the collimator is deployed as a high-frequency obturator. If, during a time duration lower than the integration time of the scintillation detector, a large number of photons interact with the scintillator, the energy is added and the spectrometer records this energy as the contribution of a high energy single photon. Experimental results underlines a tangible proof of concept for a selfcalibration device, which will be valuable for the correction of the temperature drift in the scintillator-photomultiplier pair.

CEA LIST is working of the development of a high-sensitivity thermal neutron counter, based on the insertion of gadolinium in spherical plastic scintillators. As the radiation sensor comprises one or several volumes of scintillating polymers, it is, by nature, sensitive to both neutron and gamma radiations. This property allows the development of a versatile detector, provided that it is possible to algorithmically separate the signature of both types of radiations. The technologically building brick is, moreover, to be implemented in a transportable ( 15 kg) system, handled by a first intervention task force in an exposed area. The aim of this paper is to study, both through Monte Carlo simulation and experimentally, an algorithmic architecture of the measurement chain discriminating neutron and gamma events. The presentation is subdivided in three main sections. First, we detail a comparative study between the outputs of a numerical modeling tool, developed to simulate the total response of the sensor to varying radiation sources and the results of a series of experiments carried out in a controlled environment. Second, we describe the building and numerical pre-validation of a simulation code dedicated to the study of consolidated algorithmic architecture, namely γ – γ quasi-coincidence on two adjacent plastic scintillators. Third, we comment on an experimental study of γ – γ quasi-coincidence, Gd-based neutron detection and a comparison between the results and the estimated of the numerical study. We conclude with the presentation of first estimates of neutron sensitivity), gamma-background vulnerability, and n/ γ discrimination ratio for the scaled and assessed, material and algorithmic architectures.