Search

Journal of radioanalytical and nuclear chemistry (2021). doi:10.1007/s10967-021-08102-2, Published by Springer Science + Business Media B.V., Dordrecht [u.a.]

The objective of this study was to demonstrate the feasibility of localising and detecting hydrate plugs inside submarine pipelines, in situ and contactless, using neutron-induced analytical techniques. Cold and fast neutron-beam instruments at Heinz Maier-Leibnitz Zentrum (MLZ) were used to show that neutrons penetrate through the thick wall and the insulation of such pipes and even the induced gamma radiation can be detected outside to perform a non-destructive chemical analysis within the pipe. It was found that the change in the hydrogen concentration caused by a possible hydrate plug can be detected in seconds; while with a detailed analysis at a given spot lasting for a few hours, it is possible to unambiguously identify the hydrate phase inside the hydrocarbon phase.

Compact accelerator-driven neutron sources allow to operate multiple optimised target-moderator-reflector (TMR) units adapted to the requirements of the respective instruments. The compact design of the TMR units allows an efficient coupling of neutron production, neutron moderation and extraction, but requires a novel way of optimisation. The neutronic performance of different TMR units based on polyethylene, heavy water and a mixture of heavy and light water moderators together with Pb and Be reflectors and a borated polyethylene absorber is discussed. Extraction channels for thermal and cold neutrons are investigated regarding the energy and time spectra.

Applied physics / A 127(8), 576 (2021). doi:10.1007/s00339-021-04713-4, Published by Springer, New York

With the construction of the ESS, the European neutron user community is eagerly awaiting the commissioning of the brightest neutron source worldwide in 2021. Parallel to this, there is however the ongoing development of neutron science being undertaken at a dwindling number of neutron facilities worldwide. The Julich Centre for Neutron Science has started a project to develop and design compact accelerator-driven high brilliance neutron sources as an efficient and cost effective alternative to the current low- and medium-flux reactor and spallation sources with the potential to offer science and industry access to neutrons. The project aims to deliver a high brilliance neutron source (HBS), consisting of a compact neutron production and moderator system which provides thermal and cold neutrons with high brilliance efficiently extracted in an optimized neutron transport system. By shaping the experiment holistically from the source to the detector, neutron experiments could be set-up for specific scientific requirements in a flexible and efficient way for the neutron user.

[8th International Meeting of the Union for Compact Accelerator-driven Neutron Sources, UCANS-8, 2019-07-08 - 2019-07-10, Paris, France] 8th International Meeting of the Union for Compact Accelerator-driven Neutron Sources, UCANS-8, Paris, France, 8 Jul 2019 - 10 Jul 2019; Les Ulis : EDP Sciences, The European physical journal / Web of Conferences Web of Conferences : proceedings proceedings 231, 03006 pp. (2020). doi:10.1051/epjconf/202023103006, Published by EDP Sciences, Les Ulis

Journal of radioanalytical and nuclear chemistry 325(2), 641-655 (2020). doi:10.1007/s10967-020-07271-w, Published by Springer Science + Business Media B.V, Dordrecht [u.a.]

The High-Brilliance Neutron Source (HBS) project aims to design a scalable compact accelerator driven neutron source (CANS) which is competitive and cost-efficient. The concept allows one to optimize the whole facility including accelerator, target, moderators and neutron optics to the demands of individual neutron instruments. Particle type, energy, timing, and pulse structure of the accelerator are fully defined by the requirements of a given neutron instrument. In the following, we present the current status of the HBS project., Proceedings of the 14th Int. Conf. on Heavy Ion Accelerator Technology, HIAT2018, Lanzhou, China

The neutron yield of a compact accelerator driven neutron source depends strongly on the target performance. This performance is influenced by the target composition and geometry, the cooling system design and which primary particles are used. We show that the temperature difference inside the target depends directly on the target thickness determined by the ion stopping range and therefore on the type and energy of the primary particle. Deuterons with a larger stopping power show a smaller temperature difference inside the target than protons allowing thus for a better target cooling.

An analytical method for the a priori calculation of the thermal neutron die-way time was developed in order to optimize the detection of prompt and delayed gamma rays induced in the neutron activation of large samples irradiated in the facility MEDINA. The method was tested against experimental data collected in measurements of large homogeneous and heterogeneous samples using two different ways of calculating the neutron buckling value. The overall results are very satisfactory considering the simplifications performed to derive an analytical representation of the thermal neutron die-away time; the relative deviations for the experimental values are about 20%. However, when an effective buckling value is used the analytical die-away times show deviations not higher the 4% highlighting how the neutron buckling is the main source of error.

International audience; The measurement of delayed gamma rays following neutron-induced fission is simulated with MCNP 6.1 to investigate the feasibility of fissile material detection in long-lived, medium activity radioactive waste in 870 L drums. The signal from homogeneously distributed fissile material in the drum is several hundred counts in the main delayed gamma peaks of interest. In a peripheral position or in the drum center, the signal is however too small to allow for a reliable measurement

A method for the analysis of 200 L steel drums filled with various amounts of concrete and polyethylene by large sample cyclic neutron activation analysis (LS-CNAA) using a pulsed 14 MeV neutron source was developed. The elemental composition obtained for the homogenous samples was found to agree well with the expected values, the differences lying between −3 and +15%. For the heterogeneous samples, the results were found to agree with the expected values within ±39%. Depending on the polyethylene content of the samples, detection limits ranging between 14 and 24 mg kg−1 for cadmium, 520 and 740 mg kg−1 for mercury and 5.5 and 53 g kg−1 for lead were achieved for a counting time of about 30 min.

International audience; A gamma and neutron shielding for the HPGe-detector of the MEDINA facility is designed by means of MCNP5 simulations. The performance of the shielding is validated comparing MCNP5 calculations and experimental background reduction factors for easy-to-implement shielding configurations. A good correspondence between simulations and experiment is observed, which comfort the expected background reduction factors of about 4 to 5 predicted by simulation for the neutron and photon currents entering in the detector, respectively, in the case of a complete implementation of the neutron and gamma shielding in MEDINA.

In order to determine nonradioactive substances in radioactive waste packages an innovative method based on prompt gamma neutron activation analysis was developed. The methodology was verified through the analysis of a 200 L steel drum homogeneously filled with concrete. The results obtained for the elemental composition of concrete are in good agreement with the expected values, the differences lying between −9 and +5 %. The limits of detection for the toxic elements cadmium and mercury are 4 and 145 mg/kg, respectively for an assay time of 2000s.

The neutron moderating effect of concrete and polyethylene mixtures on the determination of the sample composition was investigated by irradiating 200 L steels drums with 14 MeV neutrons at the MEDINA (Multi Element Determination based on Instrumental Neutron Activation) facility. The elemental composition was evaluated using a quantification model validated for a concrete matrix in a previous work. The results obtained were found to agree with the expected values within ±34 % which is reasonable with regards to the samples heterogeneity.

The FaNGaS instrument was set up at the research reactor FRM II, Garching. It performs measurements of ( n,n ′ γ ) reactions based on the γ-ray spectroscopy technique within the fission neutrons energy region. The optimization studies of the experimental setup were carried out using MCNP simulations. A full characterization of the detector system and the neutron beam is presented. The determination of the neutron energy spectrum was done using the foil activation method. An average neutron energy of about 2 MeV and a neutron flux of 10 8 cm − 2 · s − 1 were measured.

[8th International Meeting of Union for Compact Accelerator-Driven Neutron Sources, UCANS-8, 2019-07-08 - 2019-07-10, Paris, France] 8th International Meeting of Union for Compact Accelerator-Driven Neutron Sources, UCANS-8, Paris, France, 8 Jul 2019 - 10 Jul 2019; Les Ulis : EDP Sciences, The European physical journal / Web of Conferences Web of Conferences : proceedings proceedings 231, 02002 pp. (2020). doi:10.1051/epjconf/202023102002, Published by EDP Sciences, Les Ulis

International Conference on Neutron Optics, NOP2017, Nara, Japan, 5 Jul 2017 - 8 Jul 2017; JPS conference proceedings 22, 011025 (2018). doi:10.7566/JPSCP.22.011025 special issue: "Proceedings of the International Conference on Neutron Optics (NOP2017)", Published by The Physical Society of Japan, Tokyo

In order to measure delayed gamma-rays without appreciable interferences of prompt gamma-rays, thermal neutron-die-away times in different large samples irradiated with 14 MeV pulsed neutrons at the MEDINA facility are studied. The die-away times are found to range between 2 and 5 ms according to the irradiated materials. For the acquisition of the delayed gamma-rays spectra a waiting time of about 12 ms after the end of the neutron pulses is needed.

The need for accurate nuclear reaction data of actinides is well documented and several initiatives from international organizations for improvement have been initiated in the past. This need, particularly in view of method development for non-destructive assay of nuclear waste, has generated a joint effort to use prompt and de- layed neutron activation techniques to enhance nuclear capture data of some long lived actinides such as 237 Np, 242 Pu and 241 Am in the frame of a multilateral cooperation. This research initiative is targeted to lay grounds for the development of a non-destructive active neutron interro- gation technique to quantify actinides in mixed waste and residues from decommissioning of nuclear installations for safe treatment and storage of such materials.

Segmented γ -scanning (SGS) is the most widely applied technique to non-destructively reconstruct the isotope specific activity content in radioactive waste drums. Around 25% of all radioactive waste drums at the Forschungszentrum Julich exhibit a homogeneous matrix distribution and activity distribution. Due to these conditions, a numerical method is derived, which largely enhances the accuracy and reliability of the activity reconstruction in comparison to the common analytical method by avoiding various simplifications in the calculations. The numerical method is validated with GEANT4 simulations of reference drums containing the isotopes 137 Cs and 60 Co. The activities, which are reconstructed with the numerical method in waste matrices with a density range from 0.5 g cm −3 to 2.3 g cm −3 show systematic deviations of around −2.1% and 4.0% in case of 137 Cs and 60 Co, respectively. These results show a significant improvement in accuracy compared to the analytical method revealing systematic deviations of around 14.8% and 23.2%, respectively. These results are confirmed by the reconstruction of the activity content in a reference drum from SGS measurements. Uncertainties calculated in the numerical method are halved in comparison to the analytical method. Thus, the reliability of the reconstructed activities is enhanced.

International audience; A MCNPX model of a neutron activation facility for radioactive waste characterization was built to support experimental studies. To validate this model, small samples of different elemental composition were irradiated and experimental data are compared to simulation results obtained using different neutron capture prompt gamma ray databases. The different databases employed lead to similar results. Simulated and measured net peak areas agree within ±20 %.

A new instrument for fast neutron gamma spectroscopy has been installed and tested at the Forschungsneutronenquelle Heinz Maier-Leibnitz of the MLZ in Garching. A beam with the flux of about 108 cm−2 s−1 fission neutrons is available to irradiate small samples to study inelastic neutron scattering reactions in materials. A number of relevant elements have been investigated and new gamma lines were identified in comparison to existing data libraries. The method seems to be linear with respect to measurement time and sample mass. A particular advantage of fast neutron PGAA compared to thermal neutron PGAA appears in the analysis of materials with high thermal capture cross sections.

The performance of a borehole-logging system, based on prompt-gamma-neutron-activation-analysis (PGNAA), for explosive detection was studied by Monte-Carlo simulations. The prompt gamma of nitrogen, which is a constituent of common explosive, was used to identify the unexploded ordnance (UXO). Our results show that the minimum counting time depends on the soil moisture, the cladding thickness and the explosive composition. In conjunction with the standard detection by magnetometry, the PGNAA is a promising analytical technique for definitive identification of deep buried UXOs.

Several samples of 237Np and 242Pu were irradiated in the guided cold neutron beam of the prompt gamma activation analysis facility at the Budapest Research Reactor. The thermal neutron capture cross sections of 237 Np and 242Pu were evaluated from the obtained prompt and delayed gamma ray data. The thermal neutron capture cross sections for 237Np(n,γ)238Np was found to be \( \sigma_{0} ({}^{237}{\text{Np}}) = 170.4 \pm 7.4\,{\text{b}} \) and for 242Pu(n,γ)243Pu to be \( \sigma_{0} ({}^{242}{\text{Pu}}) = 19.6 \pm 3.9\,{\text{b}} \).

Radioactive waste must be characterized in order to verify its conformance with national regulations for intermediate storage or its disposal. Segmented gamma scanning (SGS) is a most widely applied non-destructive analytical technique for the characterization of radioactive waste drums. The isotope specific activity content is generally calculated assuming a homogeneous matrix and activity distribution for each measured drum segment. However, real radioactive waste drums exhibit non-uniform isotope and density distributions most affecting the reliability and accuracy of activities reconstruction in SGS. The presence of internal shielding structures in the waste drum contributes generally to a strong underestimation of the activity and this in particular for radioactive sources emitting low energy gamma-rays independently of their spatial distribution. In this work we present an improved method to quantify the activity of spatially concentrated gamma-emitting isotopes (point sources or hot spots) in heterogeneous waste drums with internal shielding structures. The isotope activity is reconstructed by numerical simulations and fits of the angular dependent count rate distribution recorded during the drum rotation in SGS using an analytical expression derived from a geometric model. First results of the improved method and enhancements of this method are shown and are compared to each other as well as to the conventional method which assumes a homogeneous matrix and activity distribution. It is shown that the new model improves the accuracy and the reliability of the activity reconstruction in SGS and that the presented algorithm is suitable with respect to the framework requirement of industrial application.

Improved methods for the reconstruction of the isotope specific activity content in nuclear waste drums with data obtained by a gamma scanning system developed at Shanghai Jiao Tong University require an analytical function of the detector response. In this work we derive an analytical detector response function for a collimated HPGe detector with a square collimation window. The model is based on a purely geometric model respecting the configuration of the collimated detector system, the positions of radioactive point sources and the absorption of γ-rays in the matrix as well as in the HPGe crystal. We show that the derived analytical detector response function is in good agreement with data simulated by MCNP5.

This work improves the reliability and accuracy in the reconstruction of the total isotope activity content in heterogeneous nuclear waste drums containing point sources. The method is based on χ 2 ‐ fits of the angular dependent count rate distribution measured during a drum rotation in segmented gamma scanning. A new description of the analytical calculation of the angular count rate distribution is introduced based on a more precise model of the collimated detector. The new description is validated and compared to the old description using MCNP5 simulations of angular dependent count rate distributions of Co-60 and Cs-137 point sources. It is shown that the new model describes the angular dependent count rate distribution significantly more accurate compared to the old model. Hence, the reconstruction of the activity is more accurate and the errors are considerably reduced that lead to more reliable results. Furthermore, the results are compared to the conventional reconstruction method assuming a homogeneous matrix and activity distribution.

The FaNGaS instrument has been developed and constructed at the Forschungszentrum Jülich GmbH for investigation of neutron inelastic scattering reactions using the fission neutron beam SR10 at the Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II) operated by the Technische Universität München in Garching. Prompt emitted gamma rays from excited states of irradiated elements can be used for analytical purposes.

The applicability of LaBr3-detectors for the non destructive characterization of radioactive waste drums is evaluated by comparing their performance with those of HPGe-detectors. It is shown, that relevant easy-to-measure key-nuclides are identified as well as quantified with a LaBr3-detector even in complex spectra at nearly the same level of precision and reliability than with HPGe-detectors.

In view of long-term radioactive waste repository, the Differential Die away Technique (DDT) can be used to determine the amount of fissile materials, mainly 239Pu and 235U, in 225 L bituminized waste drums by detecting prompt and delayed fission neutrons between the pulses of a deuterium-tritium neutron generator. A complementary approach studied in this paper is to detect fission delayed gamma rays. Their yields were first measured with bare samples of 235U and 239Pu in REGAIN, a facility dedicated to Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the 225 L bituminized waste drums is then assessed using the MCNPX model of MEDINA, another PGNAA cell dedicated to 200 L drums at FZJ, Germany. The performances in the bituminized drum are strongly limited by the high gamma count rate due to 137Cs, which imposes the use of collimator and shield, to avoid electronics saturation, and makes the detection of the weak delayed gamma signal impossible. However, promising performances in lower-activity cemented waste drums produced in Germany are currently being studied.

Radioactive waste has to undergo a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Within the quality checking of radioactive waste packages non-destructive assays are required to characterize their radio-toxic and chemo-toxic contents. The Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety of the Forschungszentrum Jülich develops in the framework of cooperation nondestructive analytical techniques for the routine characterization of radioactive waste packages at industrial-scale. During the phase of research and development Monte Carlo techniques are used to simulate the transport of particle, especially photons, electrons and neutrons, through matter and to obtain the response of detection systems. The radiological characterization of low and intermediate level radioactive waste drums is performed by segmented γ-scanning (SGS). To precisely and accurately reconstruct the isotope specific activity content in waste drums by SGS measurement, an innovative method called SGSreco was developed. The Geant4 code was used to simulate the response of the collimated detection system for waste drums with different activity and matrix configurations. These simulations allow a far more detailed optimization, validation and benchmark of SGSreco, since the construction of test drums covering a broad range of activity and matrix properties is time consuming and cost intensive. The MEDINA (Multi Element Detection based on Instrumental Neutron Activation) test facility was developed to identify and quantify non-radioactive elements and substances in radioactive waste drums. MEDINA is based on prompt and delayed gamma neutron activation analysis (P&DGNAA) using a 14 MeV neutron generator. MCNP simulations were carried out to study the response of the MEDINA facility in terms of gamma spectra, time dependence of the neutron energy spectrum, neutron flux distribution. The validation of the measurements simulations with Mont-Carlo transport codes for the design, optimization and data analysis of further P&DGNAA facilities is performed in collaboration with LMN CEA Cadarache. The performance of the prompt gamma neutron activation analysis (PGNAA) for the nondestructive determination of actinides in small samples is investigated. The quantitative determination of actinides relies on the precise knowledge of partial neutron capture cross sections. Up to today these cross sections are not very accurate for analytical purpose. The goal of the TANDEM (Trans-uranium Actinides’ Nuclear Data – Evaluation and Measurement) Collaboration is the evaluation of these cross sections. Cross sections are measured using prompt gamma activation analysis facilities in Budapest and Munich. Geant4 is used to optimally design the detection system with Compton suppression. Furthermore, for the evaluation of the cross sections it is strongly needed to correct the results to the self-attenuation of the prompt gammas within the sample. In the framework of cooperation RWTH Aachen University, Forschungszentrum Jülich and the Siemens AG will study the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA). The system is based on a 14 MeV neutron source and an advanced detector system (a-Si flat panel) linked to an exclusive converter/scintillator for fast neutrons. For shielding and radioprotection studies the codes MCNPX and Geant4 were used. The two codes were benchmarked in processing time and accuracy in the neutron and gamma fluxes. Also the detector response was simulated with Geant4 to optimize components of the system.

The French Alternative Energies and Atomic Energy Commission (CEA Cadarache), the Forschungszentrum Jülich GmbH (FZJ), and the RWTH Aachen University (RWTH) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The design of an optimized measurement system and the assessment of its performances for realistic scenarios can be conveniently studied by numerical Monte Carlo simulation, provided the model and nuclear data offer a sufficient precision. Previous studies performed with MCNP have shown that when the nuclear data libraries lack of precision, relevant results can still be obtained by performing calculations in multiple steps (by first determining the radiative capture rate, and transporting the induced gamma toward the detector) and by injecting valid gamma-ray production data in-between [1]. In such cases, it is interesting to compare the results obtained with different codes. In the present paper, we propose to compare the MCNP and TRIPOLI codes with measurements obtained in MEDINA (Multi Element Detection based on Instrumental Neutron Activation), which is the new FZJ PGNAA facility [2]. The aim of the measurement campaign is to assess capture gamma rays of toxic elements that can be found in 200 L waste drums which are expected for geological repository.

A new analytical facility called MEDINA (Multi Element Determination based on Instrumental Neutron Activation) and based on Prompt- and Delayed-Gamma-Neutron-Activation-Analysis (P&DGNAA) using a 14 MeV neutron generator is developed for the determination of non-radioactive elements and substances in 200-l radioactive waste drums. The qualitative performance of MEDINA is studied by measuring (1) the gamma-ray spectra for the empty chamber, (2) for an empty 200-l steel drum in presence or not of 60Co and 137Cs sources and (3) for a 200-l steel drum filled with concrete in order to determine the level of interferences induced by activation products of the facility components and to investigate further optimization steps to improve the measurements sensitivity. The measurement carried out for the empty drum with the radioactive sources show the possibility to characterize radioactive waste packages containing both 60Co and 137Cs with activities ranging between 20 and 80 MBq at least.

The Forschungszentrum Julich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA) [1]. The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.

The French Alternative Energies and Atomic Energy Commission (CEA) and National Radioactive Waste Management Agency (ANDRA) are conducting an R&D program to improve the characterization of long-lived and medium activity (LL–MA) radioactive waste packages. In particular, the amount of toxic elements present in radioactive waste packages must be assessed before they can be accepted in repository facilities in order to avoid pollution of underground water reserves. To this aim, the Nuclear Measurement Laboratory of CEA-Cadarache has started to study the performances of Prompt Gamma Neutron Activation Analysis (PGNAA) for elements showing large capture cross sections such as mercury, cadmium, boron, and chromium. This paper reports a comparison between Monte Carlo calculations performed with the MCNPX computer code using the ENDF/B-VII.0 library and experimental gamma rays measured in the REGAIN PGNAA cell with small samples of nickel, lead, cadmium, arsenic, antimony, chromium, magnesium, zinc, boron, and lithium to verify the validity of a numerical model and gamma-ray production data. The measurement of a ∼20 kg test sample of concrete containing toxic elements has also been performed, in collaboration with Forschungszentrum Julich, to validate the model in view of future performance studies for dense and large LL–MA waste packages.

Nuclear activities produce radioactive wastes classified following their radioactive level and decay time. An accurate characterization is necessary for efficient classification and management. Medium and high level wastes containing long lived radioactive isotopes will be stored in deep geological storage for hundreds of thousands years. At the end of this period, it is essential to ensure that the wastes do not represent any risk for humans and environment, not only from radioactive point of view, but also from stable toxic chemicals. This PhD thesis concerns the characterization of toxic chemicals and nuclear material in radioactive waste, by using neutron activation analysis, in the frame of collaboration between the Nuclear Measurement Laboratory of CEA Cadarache, France, and the Institute of Nuclear Waste Management and Reactor Safety of the research center, FZJ (Forschungszentrum Jülich GmbH), Germany. The first study is about the validation of the numerical model of the neutron activation cell MEDINA (FZJ), using MCNP Monte Carlo transport code. Simulations and measurements of prompt capture gamma rays from small samples measured in MEDINA have been compared for a number of elements of interest (beryllium, aluminum, chlorine, copper, selenium, strontium, and tantalum). The comparison was performed using different nuclear databases, resulting in satisfactory agreement and validating simulation in view of following studies. Then, the feasibility of fission delayed gamma-ray measurements of 239Pu and 235U in 225 L waste drums has been studied, considering bituminized or concrete matrixes representative of wastes produced in France and Germany. The delayed gamma emission yields were first determined from uranium and plutonium metallic samples measurements in REGAIN, the neutron activation cell of LMN, showing satisfactory consistency with published data. The useful delayed gamma signals of 239Pu and 235U, homogeneously distributed in the 225 L matrixes, were then determined by MCNP simulations using MEDINA numerical model. Weak signals of about one hundred counts per gram of 239Pu or 235U after 7200 s irradiation were obtained. Because of the high gamma emission in the bituminized waste produced in France (about 1 TBq of 137Cs per drum), the use of collimator and/or shielding is mandatory to avoid electronic saturation, making fission delayed gamma rays undetectable. However, German concrete drums being of lower activity, their corresponding active background was measured in MEDINA with a concrete mock-up, leading to detection limits between 10 and 290 g of 235U or239Pu, depending on the delayed gamma line. In order to improve these performances, the shielding of MEDINA germanium detector was optimized using MCNP calculations, resulting in gamma and neutron background reduction factors of 4 and 5, respectively. The experimental validation of the shielding efficiency was performed by implementing easy-to-build configurations in MEDINA, which confirmed the expected background reduction factors predicted by MCNP. Thanks to an optimized detector shielding, it will also be possible to use a higher neutron emission source, like a high flux neutron generator or an electron LINAC with appropriate conversion targets, in view to further reduce detection limits., Les activités nucléaires génèrent des déchets radioactifs classés selon leur niveau d’activité et la durée de vie des radioéléments présents. La garantie d’un classement et d’une gestion optimale nécessite une caractérisation précise. Les déchets de moyenne et haute activité, contenant des radioéléments à vie très longue, seront stockés en profondeur pendant plusieurs centaines de milliers d’années, à l’issue desquelles il est nécessaire de pouvoir garantir l’absence de risques pour l’homme et l’environnement, non seulement sur le plan radiologique, mais aussi en ce qui concerne des éléments stables, toxiques du point de vue chimique. Cette thèse concerne la caractérisation par activation neutronique de ces éléments toxiques, ainsi que celle des matières nucléaires présentes dans les colis. Elle a été réalisée dans le cadre d’une collaboration entre le Laboratoire de Mesures Nucléaires du CEA Cadarache, en France, et l’institut de Gestion des Déchets Radioactifs et de Sûreté des Réacteurs du centre de recherche FZJ (Forschungszentrum Jülich), en Allemagne. La première étude a consisté à valider le modèle numérique de la cellule d’activation neutronique MEDINA (FZJ) avec le code de transport Monte Carlo MCNP. Les rayonnements gamma prompts de capture radiative d’échantillons contenant des éléments d’intérêt (béryllium, aluminium, chlore, cuivre, sélénium, strontium et tantale) ont été mesurés et comparés aux simulations avec diverses bases de données nucléaires, permettant d’aboutir à un accord satisfaisant et validant le schéma de calcul en vue des études suivantes. Ensuite, la mesure des rayonnements gamma retardés de fissions induites sur les isotopes 235U et 239Pu a été étudiée pour des fûts de 225 L contenant des enrobés bitumineux ou une matrice béton, représentatifs de déchets produits en France et en Allemagne. Les rendements d’émission des rayonnements gamma retardés de fission d’intérêt, cohérents avec ceux publiés dans la littérature, ont été déterminés à partir des mesures d’échantillons métalliques d’uranium et de plutonium dans la cellule d’activation neutronique REGAIN du LMN. Le signal utile a ensuite été extrapolé par simulation MCNP pour une répartition homogène d’isotopes 239Pu ou 235U dans les matrices considérées, en utilisant le modèle numérique de MEDINA. Des signaux faibles, de l’ordre de 100 coups par gramme d’isotope 239Pu ou 235U, ont été obtenus. Pour le colis d’enrobés bitumineux, le niveau d’irradiation gamma très élevé, dû à une activité en 137Cs de l’ordre de 1 TBq par fût, nécessiterait l’utilisation d’une collimation et/ou d’écrans pour éviter la saturation de l’électronique de mesure, rendant indétectables les rayonnements gamma retardés de fission. Les colis de déchets bétonnés produits en Allemagne présentant un niveau d’activité plus faible, il a été possible d’estimer des limites de détection allant de 10 à 290 g d’isotope fissile 235U ou 239Pu, selon la raie gamma considérée, suite à la mesure du bruit de fond actif dans MEDINA avec une matrice béton maquette. Afin d’améliorer ces performances, le blindage du détecteur germanium de MEDINA a été optimisé à l’aide de simulations MCNP, montrant la possibilité de réduire les bruits de fond gamma et neutron d’un facteur 4 et 5, respectivement. La validation expérimentale de l’efficacité du blindage a été effectuée à partir de configuration simples à implémenter dans MEDINA, confirmant les facteurs de réduction attendus. Un blindage du détecteur optimal permettrait d’améliorer les limites de détection et aussi d’utiliser une source de neutrons d’intensité supérieure, comme un générateur de neutron à haut flux ou un accélérateur linéaire d’électrons avec une cible de conversion appropriée.