Nuclear power plants provide about 11 percent of the world's electricity production. For fission process is uranium fuels used with varying percentage of enrichment 235U for most of nuclear reactors. Uranium reserves are reducing and their mining cost increases. Therefore, the thorium fuel is discussed as revolution fuel for current and future nuclear power plants. This diploma thesis deals with possibility of thorium fuel utilization at various types of nuclear reactors with a focus on light water reactors. The practical part of the thesis is focused on simulation and calculations of various uranium dioxide and thorium dioxide layers at the fuel rods. Model of WWER 440 reactor was developed for the calculations with the addition of thorium fuel. The model simulates burning out of fuel for 5 years, with monitoring of fuel behavior and tracking changes of each material. The thesis tries to define the suitable ratio and parameters of layers combination of uranium and thorium fuel. For these ratios and parameters the thesis tries to give sufficient amount of computational analyzes.

It is important to have a method of experimentally calculating fission product yields. Statistical calculations and simulations produce very large uncertainties. Experimental calculations, depending on the methods used, tend to produce lower uncertainties. This work set up a method to calculate fission product yields using gamma ray spectroscopy. In order to produce a method that was theoretically sound, a simulation was set up using OrigenArp to calculate theoretical concentrations of fission products from the irradiation of natural uranium. From these concentrations, the fission product yields were calculated to verify that they would agree with expected values. Moving forward in the work, the total flux at the point of irradiation, in the pneumatic transfer system, was calculated and determined to be 3.9070E+11 ± 6.9570E+10 n/cm^2/s at 100 kW. Once the flux was calculated, the method for calculating fission product yields was implemented and yields were calculated for 10 fission products. The yields calculated were in very good agreement (within 10.04%) with expected values taken from the ENDF-349 library. This method has strong potential in nuclear forensics as it can provide a means for developing a library of experimentally-determined fission product yields, as well as rapid post-nuclear detonation analysis. / text

Nowadays we are all witnesses of a competition of facilities at different countries to study unknown regions of neutron rich nuclei. Much efforts are devoted to understand the role of neutron excess and its influence on nuclei in vicinity of closed neutron shells. One of the means to investigate nuclear structure is in beta-decay. Once a nucleus is proven to exist, its beta-decay properties, such as T1/2 and Pn (probability of beta-delayed neutron emission), which are relatively easy to measure, can provide the first hints on the nuclear structure. On the r-process site, "waiting points"(nuclei on closed neutron shells) has significant effects on the r-process dynamics and the abundance distribution. The actual side and the astrophysical conditions under which the nuclear synthesis takes place are still not certainly known - since r-process nuclei are difficult to produce and to study experimentally, input parameters for r-process calculations are mostly derived from theoretical models. As it has been seen lately, most of the theories have failed to reproduce newly measured data sets near shell closures. With new experimental data already (or shortly) available theoretical approaches can be adjusted. Since a beta-delayed neutron emission becomes strong if not dominating decaying channel for nuclei far stability, a proper neutron detector to study their properties is indispensable. To conduct the appropriate investigations, in the frame of the present thesis, in close collaboration with JINR (Dubna) a new detection system was constructed. It consists of 80 ³He-filled counters, 4π beta detector and a HPGe in order to measure simultaneously beta, gamma, neutron activity. The development of such a detection system system, currently installed at ALTO ISOL facility, was the first objective of the thesis. Then, during two experimental campaigns conducted to investigate beta decay properties of neutron rich nuclei in the neighborhood of N=50, N=82 the workability of the newly produced detection system was proven. In the vicinity of ⁷⁸Ni: half- lives and probability of beta-delayed neutron emission for ⁸º,⁸²,⁸³,⁸⁴Ga were measured. We were the first to observe the structure of ⁸¹,⁸² Ge via beta neutron gated gamma spectra. Thanks to the neutron detection channel the absolute intensities of beta decay were proposed for the first time. In the vicinity of ¹³²Sn the half lives of ¹²³Ag, ¹²⁴Ag, ¹²⁵Ag and ¹²⁷In, ¹²⁸In was measured. For the first time the beta delayed neutron emission was observed for ¹²⁶Cd, its Pn value also measured. Based on the data obtained we come to the conclusion that to figure out the relative contribution of allowed and forbidden decays more theoretical efforts should be done crossing the N=50 shell. Whereas in the vicinity of N=82 shell more experimental challenge are required.

Le niveau de conversion des réacteurs CANDU et REP en cycle thorium a été étudié dans l'optique d'une utilisation en troisième et dernière strate de scénarios symbiotiques. Le plutonium du combustible REP usé serait par exemple utilisé en CANDU Th/Pu pour produire de l'233U, qui alimenterait ces réacteurs à eau et haute conversion. En cas d'augmentation importante de la production d'énergie à partir d'uranium, cette alternative basée sur des réacteurs existants pourrait suppléer une IVe génération trop tardive. Pour évaluer la compétitivité de tels scénarios, des calculs de cycles détaillés ont été effectués selon une méthodologie de simulation de coeur développée pour le CANDU-6 et adaptée au REP de type N4. Le CANDU Th/233U enrichi à 1.30 wt% est régénérateur, avec un burnup court de 7 GWj/t. Augmenter légèrement l'enrichissement allonge considérablement le cycle, au prix d'une sous-génération. Multirecycler conduit également à une perte de conversion, qui peut néanmoins être compensée par un chargement fissile hétérogène. La conversion à puissance standard est moins bonne en REP Th/233U qu'en CANDU (inventaire fissile réduit de moitié après 50 GWj/t) mais peut être améliorée par sous-modération. L'analyse neutronique montre que l'essentiel du gap de conversion entre CANDU et REP vient des conditions opératoires économes en neutrons du CANDU. Des scénarios ont été comparés du point de vue de l'économie d'uranium et de l'aval du cycle dans les deux cas, et ont confi rmé l'intérêt du CANDU. Deux pistes de recherche ont été identi fiées : l'évaluation de la sûreté des CANDUs au thorium par cinétique avec contre-réactions thermiques, et l'étude de coeurs fortement sous-modérés en cuve standard de REP.

The build–up of 6Li and 3He, that is, the strong thermal neutron absorbers or the so called "neutron poisons", in the beryllium reflector changes the physical characteristics of the reactor, such as reactivity, neutron spectra, neutron flux level, power distribution, etc.; furthermore,gaseous isotopes such as 3H and 4He induce swelling and embrittlement of the reflector. The SAFARI–1 research reactor, operated by Necsa at Pelindaba in South Africa, uses a beryllium reflector on three sides of the core, consisting of 19 beryllium reflector elements in total. This MTR went critical in 1965, and the original beryllium reflectors are still used. The individual neutron irradiation history of each beryllium reflector element, as well as the impact of beryllium poisoning on reactor parameters, were never well known nor investigated before. Furthermore, in the OSCAR{3 code system used in predictive neutronic calculations for SAFARI–1, beryllium reflector burn–up is not accounted for; OSCAR models the beryllium reflector as a non–burnable, 100% pure material. As a result, the poisoning phenomenon is not accounted for. Furthermore, the criteria and hence the optimum replacement time of the reflector has never been developed. This study presents detailed calculations, using MCNP, FISPACT and the OSCAR{3 code system, to quantify the influence of impurities that were originally present in the fresh beryllium reflector, the beryllium reflector poisoning phenomenon, and further goes on to propose the reflector's replacement criteria based on the calculated fluence and predicted swelling. Comparisons to experimental low power flux measurements and effects of safety parameters are also established. The study concludes that, to improve the accuracy and reliability of the predictive OSCAR code calculations, beryllium re flector burn–up should undoubtedly be incorporated in the next releases of OSCAR. Based on this study, the inclusion of the beryllium reflector burn–up chains is planned for implementation in the currently tested OSCAR–4 code system. In addition to beryllium reflector poisoning, the replacement criteria of the reflector is developed. It is however crucial that experimental measurements on the contents of 3H and 4He be conducted and thus swelling of the reflector be quantifed. In this way the calculated results could be verified and a sound replacement criteria be developed. In the absence of experimental measurements on the beryllium reflector, the analysis and quantifcation of the calculated results is reserved for future studies. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The build–up of 6Li and 3He, that is, the strong thermal neutron absorbers or the so called "neutron poisons", in the beryllium reflector changes the physical characteristics of the reactor, such as reactivity, neutron spectra, neutron flux level, power distribution, etc.; furthermore,gaseous isotopes such as 3H and 4He induce swelling and embrittlement of the reflector. The SAFARI–1 research reactor, operated by Necsa at Pelindaba in South Africa, uses a beryllium reflector on three sides of the core, consisting of 19 beryllium reflector elements in total. This MTR went critical in 1965, and the original beryllium reflectors are still used. The individual neutron irradiation history of each beryllium reflector element, as well as the impact of beryllium poisoning on reactor parameters, were never well known nor investigated before. Furthermore, in the OSCAR{3 code system used in predictive neutronic calculations for SAFARI–1, beryllium reflector burn–up is not accounted for; OSCAR models the beryllium reflector as a non–burnable, 100% pure material. As a result, the poisoning phenomenon is not accounted for. Furthermore, the criteria and hence the optimum replacement time of the reflector has never been developed. This study presents detailed calculations, using MCNP, FISPACT and the OSCAR{3 code system, to quantify the influence of impurities that were originally present in the fresh beryllium reflector, the beryllium reflector poisoning phenomenon, and further goes on to propose the reflector's replacement criteria based on the calculated fluence and predicted swelling. Comparisons to experimental low power flux measurements and effects of safety parameters are also established. The study concludes that, to improve the accuracy and reliability of the predictive OSCAR code calculations, beryllium re flector burn–up should undoubtedly be incorporated in the next releases of OSCAR. Based on this study, the inclusion of the beryllium reflector burn–up chains is planned for implementation in the currently tested OSCAR–4 code system. In addition to beryllium reflector poisoning, the replacement criteria of the reflector is developed. It is however crucial that experimental measurements on the contents of 3H and 4He be conducted and thus swelling of the reflector be quantifed. In this way the calculated results could be verified and a sound replacement criteria be developed. In the absence of experimental measurements on the beryllium reflector, the analysis and quantifcation of the calculated results is reserved for future studies. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The detonation of a radiological dispersion device (RDD) may result in a situation where many individuals are exposed to contamination due to the inhalation of radioactive materials. Assessments of contamination may need to be performed by emergency response personnel in order to triage the potentially exposed public. The feasibility of using readily available standard 2x2-inch sodium-iodide detectors to determine the committed effective dose to a patient following the inhalation of a radionuclide has been investigated. The 2x2-NaI(Tl) detector was modeled using the Monte Carlo simulation code, MCNP-5, and was validated via a series of experimental benchmark measurements using a polymethyl methacrylate (PMMA) slab phantom. Such validation was essential in reproducing an accurate detector response. Upon verification of the detector model, six anthropomorphic phantoms, based on the MIRD-V phantoms, were modeled with nuclides distributed to simulate inhaled contamination. The nuclides assessed included Am-241, Co-60, Cs-137, I-131, and Ir-192. Detectors were placed at four positions on the phantoms: anterior right torso, posterior right torso, anterior neck, and lateral left thigh. The detected count-rate varied with respect to detector position, and the optimal detector location was determined on the body. The triage threshold for contamination was set at an action level of 250-mSv of intake. Time dependent biokinetic modeling was employed to determine the source distribution and activity in the body as a function of post-inhalation time. The detector response was determined as a function of count-rate per becquerel of activity at initial intake. This was converted to count-rate per 250-mSv intake for triage use by first responders operating the detector to facilitate triage decisions of contamination level. A set of procedure sheets for use by first responders was compiled for each of the phantoms and nuclides investigated.

This diploma thesis deals with storing the spent nuclear fuel and reviewing its safety. The theoretical part analyzes the processes taking place while the fuel is being used, such as fission, isotopic changes, fission gas release, cracking, swelling and densification of fuel pellet. The thesis is also focused on handling the spent fuel and on the way it makes from the reactor, through the spent fuel pool, the transportation, various kinds of storing, till the reprocessing and final deep geological repository. Furthermore, this part of the thesis briefly discusses computing code MCNP, its main characteristics, input files and using. The practical part of the work is focused on creating the model of the spent fuel pool located next to the nuclear reactor WWER 440/V213. This type was chosen, because it is the most used type of nuclear reactor in Czech Republic and Slovakia. With the help of the code MCNP, the multiplication factor of the main configurations of the fuel in the pool was calculated, and then the required safety regulations to ensure sufficient subcriticality, so its safety, were checked. Next, several analysis were performed using this model. These analyses were concerning the temperature of coolant, fuel and the use of various nuclear data libraries. In the future this model can be used to realize new analyses with new kinds of fuels, materials and data libraries.

The master’s thesis provides an overview of available neutron sources in terms of neutron yields and energy spectrum of emitted neutrons. Reactions of neutrons with matter, especially neutron scattering and radiation capture, are described. The possibilities neutron neutron detection and spectrometry are also described. The following experiment deals with a design of suitable shielding materials and the analysis of the moderated energy spectrum of neutron flux. The properties of the neutron field were measured using detection by activation. Subsequently, a simulation of the problem was performer in the MCNP program. In the end, the achieved results are compared and evaluated.

[ES] En el apoyo a la mejora de la calidad de medida en el Laboratorio de Radiactividad Ambiental (LRA) de la Universitat Politècnica de València (UPV), los códigos de Monte Carlo representan una potente herramienta para complementar las tareas relacionadas con la medida de la radiactividad ambiental, tales como la calibración en eficiencia de detectores de semiconductor, determinación de factores de corrección por coincidencia y caracterización de dosímetros de termoluminiscencia, entre otras. En la presente Tesis se desarrollan modelos de simulación en Monte Carlo a través de códigos y herramientas como MCNP6 y GEANT4. En primer lugar, se han realizado dos modelos de detector de semiconductor para espectrometría gamma, uno tipo HPGe (High Purity Germanium) y el otro BEGe (Broad Energy Germanium), ambos de alta pureza de germanio. Ambos detectores se emplean en las actividades y procedimientos rutinarios que se realizan en el LRA-UPV. Se detalla el procedimiento de caracterización geométrica de los detectores de semiconductor, así como del volumen activo del cristal de germanio hasta obtener un modelo geométrico optimizado. Por otro lado, se ha obtenido un tercer modelo de simulación, pero en este caso de un dosímetro de termoluminiscencia, en concreto de un TLD-100 LiF:Mg,Ti, modelo que se emplea en el servicio de dosimetría personal de la UPV. En el modelo de simulación se incluye una fuente puntual colimada de Rayos-X y el fantoma recomendado por la ISO 4037-3 (water slab phantom). Se obtiene la función de respuesta del dosímetro relativa a la energía del 137Cs y se estudia su comportamiento para diferentes condiciones de irradiación (calidad del haz de Rayos-X y ángulo de incidencia) así como para diversos materiales termoluminiscentes además del LiF. Los modelos de simulación para espectrometría gamma se han utilizado principalmente para la obtención de curvas de calibración en eficiencia para diferentes geometrías y matrices de medición, así como para el cálculo de factores de corrección por pico suma tanto para las series naturales del 238U y 232Th como para radioisótopos específicos empleados en la calibración experimental de los equipos. Por otro lado, se han aplicado los modelos de simulación en el contexto de respuesta en emergencias nucleares o radiológicas. En concreto, el modelo del detector BEGe se ha utilizado para desarrollar una metodología de optimización del proceso de medición de muestras radiactivas en matrices de agua de alta actividad. Esta metodología consiste en un procedimiento logístico que incluye un cribado o screening de emergencias soportado por simulaciones Monte Carlo, enfocado en elegir la configuración óptima de medición para obtener resultados fiables y precisos minimizando la manipulación de la muestra radiactiva. De este modo se reduce el tiempo de respuesta por parte del laboratorio, así como el riesgo de contaminación y exposición a dosis. / [EN] In support of the improvement of measurement quality at the Laboratorio de Radiactividad Ambiental (LRA) of the Universitat Politècnica de València (UPV), the Monte Carlo codes represent a powerful tool to complement the tasks related to the measurement of environmental radioactivity, such as the calibration in efficiency of semiconductor detectors, determination of coincidence summing correction factors and characterization of thermoluminescence dosimeters, among others. In the present thesis, Monte Carlo simulation models are developed using the MCNP6 code and the GEANT4 toolkit. Two semiconductor detector models for gamma spectrometry have been made, one type HPGe (High Purity Germanium) and the other one a BEGe (Broad Energy Germanium), both of high purity germanium. Both detectors are used in the routine activities and procedures carried out by the LRA-UPV. The geometric characterization procedure of the semiconductor detectors is detailed, as well as the active volume of the germanium crystal until an optimized geometric model is obtained. On the other hand, a third simulation model has been developed, but in this case from a thermoluminescence dosimeter, specifically from a TLD-100 LiF:Mg,Ti, a model used in the personal dosimetry service for the monitoring and assessment of the professionally exposed workers belonging to the UPV radioactive facility. The simulation model includes a collimated X-ray point source and the phantom recommended by the ISO 4037-3 (water slab phantom). The response function of the dosimeter relative to the energy of 137Cs is obtained and its behaviour is studied for different irradiation conditions (quality of the X-ray beam and angle of incidence) as well as for several thermoluminescent materials in addition to the LiF. The simulation models for gamma spectrometry have been used mainly to obtain efficiency calibration curves for different geometries and measurement matrices and to calculate true summing correction factors for both the 238U and 232Th natural decay series and for specific radioisotopes used in the experimental calibration of the equipment. On the other hand, simulation models have been applied in the context of nuclear or radiological emergency response. Specifically, the BEGe detector model has been used to develop a methodology for optimisation of the process of measuring radioactive samples in water matrices of high activity. This methodology consists of a logistic procedure that includes a screening for emergencies. This procedure is supported by Monte Carlo simulations, focused on determining the optimal measurement configuration to obtain reliable and accurate results, minimizing the manipulation of the radioactive sample. Therefore, the response time by the laboratory is reduced, as well as the risk of contamination and dose exposure. / [CA] En el suport a la millora de la qualitat de mesura en el Laboratori de Radioactivitat Ambiental de la Universitat Politècnica de València, els codis de Monte Carlo representen una potent eina per a complementar les tasques relacionades amb la mesura de la radioactivitat ambiental, com ara el calibratge en eficiència de detectors de semiconductor, determinació de factors de correcció per coincidència i caracterització de dosímetres de termoluminescència, entre altres. En la present tesi es desenvolupen models de simulació en Monte Carlo a través de codis i eines com MCNP6 i GEANT4. En primer lloc s'han realitzat dos models de detector de semiconductor per a espectrometria gamma, un tipus HPGe (High Purity Germanium) i l'altre BEGe (Broad Energy Germanium), tots dos d'alta puresa de germani. Aquests detectors s'empren en les activitats i procediments rutinaris que es realitzen en el Laboratori de Radioactivitat Ambiental (LRA) de la Universitat Politècnica de València (UPV). Es detalla el procediment de caracterització geomètrica dels detectors de semiconductor, així com del volum actiu del cristall de germani fins a obtindre un model geomètric optimitzat. D'altra banda, s'ha obtingut un tercer model de simulació, però en aquest cas d'un dosímetre de termoluminescència, en concret d'un TLD-100 LiF:Mg,Ti, model que s'empra en el servei de dosimetria personal de la UPV. En el model de simulació s'inclou una font puntual col·limada de Raigs-X i el fantoma recomanat per l'ISO 4037-3 (water slab phantom). S'obté la funció de resposta del dosímetre relativa a l'energia del 137Cs i s'estudia el seu comportament per a diferents condicions d'irradiació (qualitat del feix de Raigs-X i angle d'incidència) així com per a diversos materials termoluminescents a més del LiF. Els models de simulació per a espectrometria gamma s'han utilitzat principalment per a l'obtenció de corbes de calibratge en eficiència per a diferents geometries i matrius de mesurament així com per al càlcul de factors de correcció per pic suma tant per a les sèries naturals del 238U i 232*Th com per a radioisòtops específics utilitzats en el calibratge experimental dels equips. D'altra banda, s'han aplicat els models de simulació en el context de resposta en emergències nuclears o radiològiques. En concret, el model del detector BEGe s'ha utilitzat per a desenvolupar una metodologia d'optimització del procés de mesurament de mostres ambientals radioactives en matrius d'aigua d'alta activitat.. Aquesta metodologia consisteix en un procediment logístic que inclou un screening o cribratge d'emergències, suportat per simulacions Monte Carlo, enfocat a triar la configuració òptima de mesurament per a obtindre resultats fiables i precisos minimitzant la manipulació de la mostra radioactiva. D'aquesta manera es redueix el temps de resposta per part del laboratori, així com el risc de contaminació i exposició a dosi. / Finalmente, a la Universitat Politècnica de València por la financiación a través de la beca de Formación de Personal Investigador (FPI)-Subprograma 2 de la convocatoria de 2015 y a la Cátedra CSN-UPV Vicente Serradell / Ordóñez Ródenas, J. (2020). Desarrollo de Modelos de Simulación por Monte Carlo como Apoyo a la Medida de Radiactividad Ambiental en Operación Rutinaria y de Emergencias [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/152188 / TESIS