Possible fuel cycle scenarios for a helium-cooled, sub-critical, fast reactor with a fusion neutron source for the transmutation of spent nuclear fuel have been analyzed. The transmutation rate was set by the 3000MWth fission power output. The primary objective was to achieve greater than 90% burn of the transuranic (TRU) fuel obtained from spent nuclear fuel. A secondary objective was to examine the possibility of achieving this deep burn without reprocessing after initial fabrication of the TRU into coated particle TRISO fuel. Four sets of 5-batch fuel cycle scenarios, differing in the constraints imposed on the beginning of cycle (BOC) k-eff and the end of cycle (EOC) neutron source strength (characterized by the fusion neutron source power level), were evaluated. In scenario A, BOC k-eff was required to be 0.95 and EOC Pfus less than 200 MWth was required. In scenario B, the restriction was removed to allow less reactive BOC fuel loadings, while the 200 MW upper limit on EOC Pfus was retained. It was found that the primary objective of greater than 90% TRU burn-up could be achieved by repeatedly reprocessing the TRISO TRU fuel particles to remove fission products and add fresh TRU makeup at the end of each 5-batch burn cycle, without needing to increase the fusion neutron source power above 100 MWth when the BOC k-eff is restricted to 0.95. The secondary objective of obviating processing could only be accomplished when the restriction was removed and recycling was employed or when both EOC Pfus and BOC k-eff restrictions were removed in a single-pass deep burn fuel cycle. In scenario C, with both the BOC k-eff limit and the fusion power limit unrestricted, greater than 90% TRU burn-up was achieved without reprocessing the TRISO TRU fuel particles, which could then be buried intact in a high-level waste repository, but a neutron source rate of 3370 MWth was required. In scenario D, with only the BOC k-eff limit unrestricted, greater than 90% TRU burn-up was achieved without reprocessing by the continuous recycle of TRISO particles through the reactor.

In view of the advantages of its type, the decision to construct a neutron source of the particle accelerator type was made. The purpose of this thesis is to survey the problems encountered in the construction of the source.

Les spectres des rayons γ prompt de fission sont des données nucléaires importantes pour la physique des réacteurs, en tant qu'entrée pour les calculs de chauffage gamma, puisque l'effet de chauffage gamma peut être sous-estimé jusqu'à ~28% avec les données nucléaires actuelles. De plus, les nouvelles informations sur les rayons γ prompts de fission seront utiles du point de vue de la physique fondamentale, où les résultats peuvent être comparés à de nombreuses prédictions théoriques concurrentes pour affiner les modèles du processus de fission. Les spectres des rayons γ prompts de fission ont été mesurés pour la fission induite par neutrons rapides de ²³⁸U et de ²³⁹Pu, en utilisant des neutrons rapides générés à partir de la source LICORNE. Le dispositif expérimental se composait d'une chambre d'ionisation et de différents types de détecteurs à scintillation, dont les détecteurs LaBr₃ et PARIS phoswich. Une procédure d'analyse, comprenant le unfolding et la simulation de la réponse aux rayons γ dans les détecteurs à scintillation, est mise au point pour extraire le spectre des rayons γ prompts de fission et les caractéristiques spectrales correspondantes. Les résultats expérimentaux sont comparés aux codes de modélisation de fission GEF et FREYA. Cette comparaison révèle que les caractéristiques spectrales sont liées aux conditions énergétiques, aux rendements isotopiques et au moment angulaire des fragments de fission. La dépendance énergétique des caractéristiques spectrales montre que l'émission des rayons γ est tout à fait insensible à l'énergie neutronique incidente. Toutefois, on observe une forte dépendance à l'égard du système fissionnant particulier. / Prompt fission γ-ray spectra are important nuclear data for reactor physics, as an input for gamma heating calculations, since the gamma heating effect can be under-estimated by up to ~28% with present nuclear data. Furthermore the new prompt fission γ-ray information will be useful from a fundamental physics point of view, where results can be compared with many competing theoretical predictions to refine models of fission process. Prompt fission γ-ray spectra have been measured for the fast-neutron-induced fission of ²³⁸U and ²³⁹PuPu, using fast neutrons generated from the LICORNE source. The experimental setup consisted of an ionization chamber and different types of scintillation detectors, including LaBr₃ and PARIS phoswich detectors. An analysis procedure, including unfolding and recovering the γ-ray response in the scintillation detectors, is developed to extract the prompt fission γ-ray spectrum and corresponding spectral characteristics. The experimental results are compared to the fission modeling codes GEF and FREYA. This comparison reveals that the spectral characteristics are related to the energetic conditions, isotopic yields and angular momentum of the fission fragments. The energy dependence of the spectral characteristics shows that the prompt γ-rays emission is quite insensitive to the incident neutron energy. However, a strong dependence on the particular fissioning system is observed.

The goal of this project was to determine the source strength of an Americium-Beryllium neutron source by irradiating metal foils of different elements. The irradiated foils were then placed on a High Purity Germanium detector and γ-decay from the decay of the neutron induced reaction products was observed and used to deduce the neutron flux. The results have been found to be in the order of magnitude of S = 2 · 105 neutrons/second, which is in fair agreement with the source strength calculated from the manufacturer’s specifications. In addition, in order to improve the reproducibility of the measurements, flexible holders have been created using 3D printers.

La méthode de mesure de référence du débit d’émission de sources neutroniques se fonde sur la technique du bain de manganèse. Elle est destinée à étalonner des sources de neutrons utilisant des radionucléides (241AmBe, 239PuBe, 252Cf,…) en termes de débit d’émission neutronique sous 4π sr. Ce dispositif est complété par un banc de mesure de l’anisotropie d’émission utilisant un support rotatif et un compteur long de type BF3. La source à mesurer est immergée dans une solution de sulfate de manganèse et les neutrons émis sont capturés par les constituants du bain. Dans une configuration classique (sphère de bain de manganèse de 1 m de diamètre et solution concentrée), environ la moitié de ces neutrons conduisent à la création de 56Mn par réaction (n, γ) sur 55Mn. Le radionucléide 56Mn a une période radioactive d’environ 2,6 heures et le bain de manganèse atteint son activité de saturation en 56Mn quand le nombre d’atomes radioactifs créés par unité de temps devient égal au nombre d’atomes se désintégrant pendant ce même temps. Le débit d’émission de la source peut alors être déduit de l’activité en 56Mn de la solution à saturation, via une modélisation ad hoc des réactions nucléaires se produisant dans le bain. Cette installation a été récemment rénovée au LNE-LNHB afin de respecter les règles de sécurité et de radioprotection en vigueur. Cette rénovation a été l’occasion de moderniser et de remettre à niveau les méthodes de mesure et de modélisation du bain et d’entreprendre une étude sur le développement d’un détecteur original pour la mesure directe en ligne de l’activité du manganèse. Ce détecteur est fondé sur la méthode de mesure par coïncidences β-γ. La voie bêta est constituée de deux photomultiplicateurs permettant de détecter l’émission de lumière due à l’effet Cerenkov et la voie gamma utilise un détecteur à scintillateur solide. L’intérêt de cette méthode de mesure est qu’elle permet d’avoir accès à l’activité du bain sans nécessiter d’étalonnage préalable, contrairement à la méthode classique qui utilise un compteur gamma et nécessite la fabrication d’une source de haute activité. Le principe de mesure a été validé à l'aide d'un prototype de détecteur et d'une modélisation effectuée à l'aide du code de calcul stochastique GEANT4. Le détecteur définitif a été réalisé et les mesures obtenues ont été comparées à celles données par une méthode primaire présente au laboratoire. Par ailleurs, des modélisations du bain de manganèse effectuées sous GEANT4, MCNPX et FLUKA, ont été comparées afin de choisir le code le plus fiable. Cette comparaison a permis d'identifier des lacunes notamment dans le code GEANT4 ainsi que des facteurs d'incertitude nécessitant une attention particulière, tels que la modélisation de l'émission neutronique et le choix des sections efficaces. Enfin, un étalonnage de source neutronique a été réalisé grâce à la méthode Cerenkov-gamma et aux facteurs correctifs donnés par la nouvelle modélisation du bain sous MCNPX. Ces mesures ont été complétées dans le cadre d'une comparaison comprenant également des mesures par l'ancienne méthode après étalonnage du couple bain/détecteur par irradiation d'une cible de manganèse en réacteur. Au terme de cette étude, plusieurs voies d'améliorations ont été proposées, dont certaines font déjà l'objet de travaux au LNHB. / The manganese bath technique is the reference method for neutron source emission rates calibration. It is used to calibrate neutron sources using radionuclides (AmBe, PuBe, 252Cf,…) in terms of neutron emission rate under 4π sr. As a complement to this technique, the anisotropy of the source is measured using a rotating source holder and a neutron long counter. The neutron source to be measured is immersed in a manganese sulphate solution whereby the emitted neutrons are captured within the bath contents. In a typical configuration (a 1m diameter sphere and a concentrated solution), approximately half of the neutrons lead to the creation of 56Mn via the 55Mn(n, γ) capture reaction. The 56Mn radionuclide has a half-life of approximately 2.6 hours and the bath reaches saturation when the number of nuclei decaying is equal to the number of nuclei created per unit time. The neutron emission rate from the source can then be deduced from the 56Mn activity at saturation, assuming proper modelling of the nuclear reactions occuring in the bath. The manganese bath facility at LNE-LNHB has been recently refurbished in order to comply with appropriate safety and radioprotection regulations. This has lead to the upgrading of both the measurement methodology and the modelling of the bath, and a study on the development of a new detector for the on-line measurement of the manganese activity was started. This new detector uses the β-γ coincidence measurement method. The bêta channel consists of two photomultipliers tubes which allow the detection of Cerenkov light, and the gamma channel uses a solid scintillation detector. The advantage of this measurement method is that it allows the determination of the bath activity without any prior calibration, unlike the former method which uses a gamma-ray detector calibrated using a high activity manganese source. The principle of the Cerenkov-gamma coincidence measurement has been validated by a prototype of the detector and via modelling of the system using the stochastic transport code GEANT4. The final detector has also been made and the results obtained have been compared to those from a primary measurement method already in use at LNE-LNHB. Furthermore, a comparison of the results from modelling the manganese bath with GEANT4, MCNPX and FLUKA have been undertaken to find the most reliable code. This comparison lead to the identification of various weaknesses, particularly in GEANT4, and several uncertainty factors, such as the modeling of the neutron emission and the choice of the cross-section library. Finally, neutron source calibration has been carried out with the Cerenkov-gamma method and the correction factors given by the new modeling of the bath using MCNPX. These results have been complemented with a comparison with the former method simultaneously undertaken, after calibration of the detector in the bath using a 56Mn source irradiated in a nuclear reactor. At the end of this study, several improvements have been proposed, from which a number are currently under development at LNE-LNHB.

Within the framework of an Institutional Partnership of the Alexander von Humboldt Foundation, the Budker Institute of Nuclear Physics Novisibirsk (BINP) and Forschungszentrum Dresden-Rossendorf worked together in a joint project devoted to the research at the coupled GDT-SHIP facility of the BINP with the focus on the study of plasma phenomena within the SHIP mirror section. The project began at July 1st, 2005 and ended on August 30th, 2008. It included work packages of significant theoretical, computational and analyzing investigations. The focus of this final report is on the presentation of results achieved whereas the work that was done is described briefly only. Chapter 2 illustrates the GDT-SHIP facility and describes shortly the planned topics of the SHIP plasma research. Chapter 3 explains the main extensions and modifications of the Integrated Transport Code System (ITCS) which were necessary for the calculations of the fast ion and neutral gas particle fields in SHIP, describes briefly the scheme of computations and presents significant results of pre-calculations from which conclusions were drawn regarding the experimental program of SHIP. In chapter 4, the theoretical and computational investigations of self-organizing processes in two-component plasmas of the GDT-SHIP device are explained and the results hitherto achieved are presented. In chapter 5, significant results of several experiments with moderate and with enhanced plasma parameters are presented and compared with computational results obtained with the ITCS. Preparing neutron measurements which are planned for neutron producing experiments with deuterium injection, Monte Carlo neutron transport calculations with the MCNP code were also carried out. The results are presented. Finally, from the results obtained within the joint research project important conclusions are drawn in chapter 6.

No description available.

Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. Experience has shown that the performance of reactivity measurement methods in terms of accuracy and applicability is highly system dependent. Further investigations are needed to increase the knowledge data bank before applying the methods to an industrial facility and to achieve license to operate such a facility. In this thesis, two systems have been subject to investigation of various reactivity measurement methods. Conditions for successful utilization of the methods are presented, based on the experimental experience. In contrast to previous studies in this field, the reactivity has not only been determined, but also monitored based on the so called beam trip methodology which is applicable also to non-zero power systems. The results of this work constitute a part of the knowledge base for the definition of a validated online reactivity monitoring methodology for facilities currently being under development in Europe (XT-ADS and EFIT). / QC 20100621

Within the framework of an Institutional Partnership of the Alexander von Humboldt Foundation, the Budker Institute of Nuclear Physics Novisibirsk (BINP) and Forschungszentrum Dresden-Rossendorf worked together in a joint project devoted to the research at the coupled GDT-SHIP facility of the BINP with the focus on the study of plasma phenomena within the SHIP mirror section. The project began at July 1st, 2005 and ended on August 30th, 2008. It included work packages of significant theoretical, computational and analyzing investigations. The focus of this final report is on the presentation of results achieved whereas the work that was done is described briefly only. Chapter 2 illustrates the GDT-SHIP facility and describes shortly the planned topics of the SHIP plasma research. Chapter 3 explains the main extensions and modifications of the Integrated Transport Code System (ITCS) which were necessary for the calculations of the fast ion and neutral gas particle fields in SHIP, describes briefly the scheme of computations and presents significant results of pre-calculations from which conclusions were drawn regarding the experimental program of SHIP. In chapter 4, the theoretical and computational investigations of self-organizing processes in two-component plasmas of the GDT-SHIP device are explained and the results hitherto achieved are presented. In chapter 5, significant results of several experiments with moderate and with enhanced plasma parameters are presented and compared with computational results obtained with the ITCS. Preparing neutron measurements which are planned for neutron producing experiments with deuterium injection, Monte Carlo neutron transport calculations with the MCNP code were also carried out. The results are presented. Finally, from the results obtained within the joint research project important conclusions are drawn in chapter 6.

Significant discrepancies had been identified by many research groups world wide between calculated and measured neutron doses from the ⁷Li(p,n) accelerator based neutron source, and therefore precise characterization of the source was needed. In this work neutron spectra from the ⁷Li(p,n) source were measured with a ³He ion chamber in the incident proton energy range from 1.95 to 2.3 MeV. The ³He detector is hypersensitive to slow neutrons, so a time-of-flight based slow neutron rejection acquisition system was built and tested. The system is based on an electrostatic proton chopper and an acquisition system working on coincidence mode. The response function of the ³He was extended down to 30 keV neutron energy and the collected neutron spectra were unfolded using two methods: van Cittert iterative algorithm with Jansson constraint, and a regularized constrained inversion. Theoretical neutron spectra emitted by the ⁷Li(p,n) source were calculated and compared with experimental unfolded spectra, as well as with results of the Monte Carlo simulations of the lithium target assembly and the walls of the experimental area. Using fluence to kerma conversion coefficients, the neutron dose was calculated and compared with results obtained from an independent experiment using the microdosimetric technique with a tissue equivalent proportional counter. Total neutron yield of the ⁷Li(p,n) reaction was measured using induced activity of ⁷Be. Results showed a negative energy offset of the incident proton beam between 50 and 58 keV with respect to the generating voltmeter indication of the accelerator terminal. Shapes of the measured neutron spectra showed significant moderation originating from neutron scattering on the lithium target assembly and walls of the experimental area. When accounting for this offset, neutron yields showed an agreement with calculated values within 22% for 1.95 MeV and within approximately 7% for higher proton energies. / Thesis / Doctor of Philosophy (PhD)