APRÈN. Portal de la producció docent del professorat de la UPC

In a typical Nuclear Power Plant (NPP), there are several safety barriers retaining the radionuclides within the NPP. A Zr-based nuclear fuel cladding constitutes the first barrier after the fuel matrix performing this crucial safety function, and many design acceptance criteria established bytheregulatorybodyareformulatedintermsofcladdingproperties. Decadesofexpertise have led to a high optimization of this technology, and exhaustive licensing processes have constantly demonstrated the capacity of Zr-based claddings to preserve its integrity in a wide range of conditions, including a spectrum of postulated accident, the so-called Design Basis Accidents (DBAs). However, harsher conditions than DBAs, like those that took place during Fukushima Daiichi's non-postulated accident, eventually cause the fuel cladding failure. Consequently, post-Fukushima international efforts have accelerated the development of Accident Tolerant Fuels(ATF),whichaimatincreasingthesafetymarginsbyensuringtheintegrityofthe f irst barrier containing the radionuclides in more severe regimes than Design Basis Accidents (DBAs). Among the proposed ATF, the Cr-coated Zr-based cladding is one of the most widely explored and most feasible technologies to be implemented in the short-term due to its costeffectiveness and higher resistance to high temperature steam environments. Before full-scale ATF implementation in Nuclear Power Plants (NPPs), a thorough safety analysis evaluation of the ATF benefits with respect to standard fuel rods is required. This task demands simulations that encompass design basis conditions and Severe Accidents, thus codes with sound capabilities in both regions of interest are needed. However, these codes often lack dedicated models that can represent new ATF designs. To contribute to this purpose and facilitate the Cr-coated ATF deployment, an oxidation model in high temperature steam environments has been developed from the available experimental data. The model relies on the evolution of three parameters, namely the Cr-coated cladding temperature, the time at which Cr oxidation starts and the growth of the chromia layer. As the first step, the model was implemented in RELAP/SCDAPSIM/MOD3.4SystemandSevereAccident code and acomplete process of verification was performed. This ensured that the model worked consistently without interfering with other simultaneous phenomena such as hydrogen production. The next step will be to validate the model with experimental data, while upcoming studies will focus on prolonged accidental sequences in the boundary between Design Extension Conditions with and without core melt, where Cr-coated ATF has a greater potential to increase the safety margins.

The reflood phase of a Loss-of-Coolant Accident (LOCA) in a nuclear power plant (NPP) is critical for safety, as it is during this phase that the peak cladding temperature (PCT) is reached. The reflood process involves rewetting the reactor core after a dry-out period, continuing until long-term cooling is achieved. NPPs must demonstrate, using thermal-hydraulic codes such as RELAP5 and TRACE, that the PCT remains within the safety limits defined by regulatory authorities. More accurate predictions of this parameter can provide additional safety margins. Therefore, a thorough understanding of the complex phenomena occurring during the reflood phase is essential. Experimental programs offer valuable data, serving as benchmarks for participants to perform simulations, evaluate codes, and develop new models and correlations. The Rod Bundle Heat Transfer (RBHT) project conducted a series of reflood experiments under specific conditions using the advanced NRC/PSURBHT facility. Aspart of this project, UPC developed a RELAP5 model that performed well but faced limitations in predicting the phenomena in the upper regions of the rod bundle. This master's thesis aims to assess TRACE's ability to accurately simulate the evolution of the reflood phase, with a particular focus on the complex post-CHF regimes present in these upper regions above the quench front. A TRACE model was developed, leveraging features such as the spacer grid model and the droplet field. The simulated results were compared with both experimental data and predictions from the RELAP5 model. A sensitivity analysis was also conducted to evaluate the impact of some parameters and physical models on the simulation outcomes, such as PCT and quench time. The results demonstrated that TRACE performs well in high reflood rate scenarios. However, limitations were observed, particularly for tests involving low flooding rates or variable flow conditions, where the model struggled to capture the heat transfer mode transitions. The sensitivity analysis revealed that the heat transfer during dispersed flow film boiling and the quench temperature significantly impacted PCT, quench times, and liquid and vapor mass balance. While TRACE shows promise in simulating certain phenomena, such as heat transfer enhancement due to the spacer grids, further development offer opportunities for improving its current capabilities.

The Breed and Burn (B&B) reactors come in handy to deal with the large stock piles of depleted uranium, using it as fuel to power the reactor. The discharge fuel of B&B contains 10% of fissile plutonium. This study analyses the possible options in generating more energy while reducing the plutonium content by using a reconditioned B&B discharge to fuel Pressurized Water Reactor (PWR). Three reprocessing processes with high proliferation resistance are considered namely, AIROX, Melt Refining and FLURO-OX process. Using AIROX processed fuel in a PWR core, no additional burnup can be achieved without compromising on the power density, and whereas when the PWR is fuelled with Melt refined fuel an additional burnup of 66 GWd/MTIHM can be achieved using the nominal power density. Melt refined fuel mixed with different types of fuels provides various options to incinerate fissile plutonium. Using smaller diameter pins an additional burnup of 95.2 GWd/MTIHM can be achieved. Using pure FLURO-OX processed fuel in a PWR core is not practical due to the positive reactivity coefficients. However with a slight modification in the FLURO-OX process uranium can be retained in the fuel, which can compensate for the positive reactivity coefficients. Fuelling PWR with such fuel an additional burnup of about 108.1 GWd/MTIHM can be achieved at the nominal power density. Instead of once through cycle, using twice burning cycle in PWR an additional burnup of about 120 GWd/MTIHM can be achieved.

To combat climate change and the effects it has and will have on the environment and human beings, it is necessary to look for alternative energy sources or other techniques to make the transition from an electricity grid dependent on fossil fuels to one free of them. In addition to wind, solar, hydroelectric and nuclear energy, nuclear fusion technology promises to combine all their advantages and become the energy source of the future. To this end, ITER, a nuclear fusion mega-reactor, is being built in France to demonstrate the technical feasibility of this type of technology. With the lessons learned at ITER, in addition to the multitude of other smaller-scale fusion reactors, it is planned to create DEMO, the first commercial nuclear fusion reactor, which will serve as a precursor to many others and usher in a new era. ITER, composed of large and complex components, uses deuterium and tritium as fuel. Deuterium is plentiful, but tritium is limited, and its consumption is increasing. Therefore, ITER and any nuclear fusion reactor that uses these hydrogen isotopes as fuel must be self-sufficient in tritium generation. To this end, modules (Test Blanket Modules) will be installed around the reactor, which will serve as a test for the generation of tritium from the also abundant lithium. The Helium Cooling System (HCS), the system under study in this project, has the function of cooling these modules, which will receive a large flow of heat from the fusion reactions that will take place in the plasma inside the reactor. As with any nuclear facility, safety takes precedence over everything else. Therefore, it is necessary to ensure that ITER and its systems are safe and meet the standards imposed by law and society. This paper shows the results obtained from the HAZOP study of the HCS, one of the many types of risk analysis that can be applied to a system. The study aims not only to show the results, but also to provide a detailed explanation of the methodology applied, so that it can be replicated in the future. Most of the consequences emerged from the deviations proposed are related to system unavailability, which are more deeply analyzed by a RAMI (Reliability, availability, maintainability, inspectability) study of the system. Other group of important consequences is that related to a compromised confinement to either workers or environment. Action items are proposed to reduce these consequences or mitigate the effects caused by them.

The following project has been conducted in the Nuclear Service of IDOM Consulting, Engineering and Architecture, within the Nuclear Safety department of their Barcelona offices. This work is in the scope of the Final Master Thesis for the Master of Nuclear Engineering of the ETSEIB (UPC). It deals with nuclear safety analysis, performed through a deterministic point of view - through the development and use of thermal hydraulics simulations - and using some tools of Probabilistic Safety Analysis (PSA). The objective of this project is to answer the concerns raised by events of flooding of nuclear power plants facilities, specifically in structures containing critical safety-related systems and components. These concerns have been partly highlighted thanks to the events of Fukushima Daiichi in 2011 when the flooding caused by the tsunami plunged the six units (three operating) of the nuclear power plant in a station blackout scenario. This work focuses on the Auxiliary Feedwater (AFW) system's building, critical in case of emergency shutdown of Pressurized Water Reactors (PWR) in order to maintain the cooling of the reactor core. The work is divided into two main sections. The first one deals with the development of a thermal hydraulics model of the AFW building of a classical 3-loops PWR, Nuclear Power Plant (NPP) in an internal flooding event from different sources. This model returns the maximum elevations of water reached through the development in time of the flood in the different areas of flooding of the building. The second part is focused on the safety analysis of the AFW system under flooding events through the use of tools of PSA like fault trees and damage equations to provide an assessment of the response of the current classical structures to this kind of events. Overall, this project aims at providing an assessment of the current response of the AFW system to flooding events and potential tracks of improvement to reach safer conditions of operation of nuclear power plant through continuous availability of critical safety-related structures, systems and components (SSC).

Benchmarking activities are crucial in the process of verification and validation of thermal-hydraulic codes, which are a powerful tool for enhancing nuclear safety. The Fukushima-Daichii disaster highlighted the necessity of strengthening nuclear power plants against Beyond Design Basis Accident scenarios, particularly Station Blackouts (SBOs), which involve the loss of both internal and external AC power sources. While conservative approaches have been successful in the control of Design Basis Accident scenarios, the refining tendencies linked to the Best-Estimate methodology are crucial in establishing design specifications for Design Extension Condition scenarios. Among the various mitigation strategies for SBO scenarios, passive safety systems emerge as significant solutions due to their independence from electrical power sources and operator intervention, relying solely on natural circulation, gravity, and stored energy. The work presented in this thesis is within the framework of the ETHARINUS joint project, and it is focused on the Test J4.2 benchmark exercise, carried on the PKL Integral Test Facility, which is based on evaluating the performance of the SACO passive system in residual heat removal during an SBO scenario. The results obtained by the UPC group using the RELAP5/MOD3.3 code, are provided with an emphasis on a thermal-hydraulic phenomenology analysis and its comparison with the experimental data.

Radiation therapy involves delivering a high dose (energy of absorbed radiation per unit mass) to a target volume within the patient. Ionizing radiation is used for penetrating the tumor tissue, causing its the death and eliminating also the cancerous cells surrounding the target. When per- formed, the aim is to minimize the impact on healthy organs by reducing the absorbed doses in their tissues. In the case of external radiation therapy, this is achieved using particle accelerators that target the affected area and direct the radiation using collimators. Before radiation therapy treatment can be administered to the patient, a specific quality control process must be conducted to ensure that the accelerator behaves appropriately according to a designed plan, from which the dose distribution to be administered to the patient's tissues has been calculated. In addition to verifying critical design parameters of the radiotherapy plan for the accelerator, measurements are taken on phantoms with detectors to verify that the admin- istered dose distribution is within acceptable tolerances. When tumors are small, the volumes that need to receive high doses are small as well, thus requiring the use of radiation beams with very small open collimator apertures. This results in the formation of small radiation fields, which present challenges for measurement, as precision dose delivery to the target is desired and they exhibit dose profiles with very abrupt changes. Therefore, detectors of small volume concentrated in a very small region are required. In this master's thesis, the behavior of two measurement devices, the SRS MapCHECK and the ArcCHECK (both from Sun Nuclear), has been studied. These devices feature semiconductor detectors and have been specifically designed to perform dose distribution measurements in pre-treatment verifications. The study was conducted by evaluating the gamma passing rates in three scenarios using the SRS MapCHECK and GAFChromic EBT-XD radiographic films (to validate the former) in the StereoPHAN phantom, as well as the ArcCHECK. Initially, square static fields ranging from 0.5 to 4 cm were irradiated using 6MV, 6MV FFF, and 10MV FFF beam energies. Subsequently, pre-treatment verifications were conducted for lung (SBRT) and brain (HSRT) stereotactic plans with target volumes having equivalent spherical diameters ranging from 3 to 1 cm. The VMAT technique was employed to deliver the radiation, which involves modulating the collimator throughout the continuous rotation of the accelerator arm. Unlike the SBRT plans, the HSRT plans were highly modulated and involved significant couch rotations. Finally, sensitivity was tested using the SEAFARER [1] methodology in a spinal SBRT plan. The results showed that the SRS MapCHECK was capable of accurately measuring fields uti- lizing VMAT techniques over volumes with equivalent diameters equal to or greater than 1.0 cm. However, the ArcCHECK encountered greater challenges in measuring all fields, showing lower sensitivity in detecting errors in radiotherapy machines with small fields. Additionally, it did not precisely measured dose distributions for equivalent diameters less than 2.0 cm, or 3.0 cm if the field was highly modulated.

Recent interest for Small Modular Reactor (SMR) and Advanced Modular Reactors (AMR) ac- celerates research and engineering activities oriented to the design of new reactor concepts, fuel cycle closure and improvement in scientific tools. In this context, code assessment and valida- tion is one of the most relevant research lines in thermal-hydraulics, as well as in reactor physics modelling; promoting benchmarks against measurements and code-to-code comparisons at in- ternational level. This master thesis is realised in the framework of an internship in the start-up newcleo, which was recently founded with the purpose of developing compact and competitive Lead Cooled Fast Reactors (LFR). In the framework of an internship realised in newcleo for the fulfilment of the Master thesis, this study aims at providing a first contribution to the applicability of ATHLET code to lead cooled fast reactors (LFRs) for both primary and secondary systems and their major components. The focus of this work has been dedicated to three different topics. Firstly, the evaluation of the robustness of liquid metal - non-condensable gas interface modelling, with benchmarking activities on two different numerical solving schemes (RELAP-SCDAPSIM and ASYST, semi- implicit; and ATHLET, implicit). Secondly, the reactor's primary and secondary loops have been modelled by means of ATHLET system code. The physical behaviour of different nodalisation options has been studied, including complex ATHLET modules, in order to achieve the most realistic conditions possible. This included the test and characterisation of ATHLET modules still undergoing validation, such as pseudo-3D modelling capabilities. Finally, the assessment of the heat transfer mechanism in a bayonet tube decay heat removal (DHR) has been carried out, including operation under different conditions. Starting from a base case DHR, optimisation of both boundary conditions and design was performed, resulting in a system capable of extracting decay heat in a safe and steady manner. It is important to note that a special effort on avoiding confidential data has been dedicated, including only open source data and calculated values derived from this data. Results obtained show the capability of the ATHLET code to cope with a liquid metal and non- condensable gas interface; while RELAP was shown to be missing some mitigation scheme to avoid numerical oscillations in physical magnitudes. This artificial oscillations may result in a loss of information about oscillations with a physical origin, if the latest is not strong enough. In addition, the interest of developing a simplified, but representative, test case of the Nuclear Steam Supply System (NSSS), for analysing code behaviour and for young engineers training purposes on a lead cooled reactor has been confirmed. A comparison of steady state variables generated with different models is included, with extra analysis regarding main parameters time evolution and steady state distribution. Finally, the validity of ATHLET was proven as a tool suitable for optimizing heat transfer mechanism and operating a bayonet tube Decay Heat Removal (DHR). Potential improvement have been proposed during the internship, mainly re- garding the loss of regenerative heat exchange between water and steam channels, outside the lead pool (cover gas region). In conclusion, the results provided confirm the interest towards ATHLET code to support Research and Development activities and training of young engineers and researchers.

In the nuclear industry all around the world, it has been built and are building hundreds of research reactors. Most of them are planned to be exploited during an extensive period of time, some of them working for more than 60 years. This fact justifies the need to establish an ageing management plan with the objective of maintaining them in good and safe operational conditions until their last exploitation instant, trying to extend their lifetime. This study objective is to establish a simple and clear methodology to constitute an ageing management plan applicable to a great diversity of research reactors. Independently of their size, design, and legislation by which they are governed. Obtaining as a result, a concise and posteriorly applicable to real life procedure. Nevertheless, it is worth noting the importance of the individualized study for each system or degradation mechanism for the different existent reactors, using this project results as a structure or base. For this purpose, it has been carried out an exhaustive study of the normative and international recommendation presented by IAEA. At the same time as the study of the materials, degradation mechanisms and most frequent systems in research reactors all around the globe. Including the explanation of the step-by-step process to develop and the different final documents of the project. In addition, is performed an objective and reproducible study of the most common key performance indicators (KPIs) to catalogue the suitability of this kind of programmes. All the previously commented theoretical methodology is applied through the study of a practical case, the research reactor SAFARI-1 in South Africa. Primarily, it includes a brief explanation of the main and specific characteristics of the reactor, in other words, its history, systems, technical aspects etc. Once the details of the reactor are determined on a qualitative way, the real application of the method for the particular case of the elements which suffer from erosion as degradation mechanism is carried out. Finalizing with the obtention of a complete programme and actually applicable for the research reactor. Finally, a study is implemented from the economic and environmental point of view of the development of the project, taking in to account the time spent and the precedence of the energetic source used during the development of this project.

Thermodiffusion, also known as the Soret effect, is a fascinating and complex phenomenon that occurs when a temperature gradient causes the migration of different components within a mixture. This phenomenon is of great interest in understanding the behaviour of nuclear fuel, as it plays a crucial role in altering concentration prof iles and, in turn, essential properties such as fuel performance, safety and lifetime. Furthermore, it has been identified as responsible for the introduction of corrosive environments for the cladding in PWRs. In this study, I explore the fundamental mechanisms governing thermodiffusion and its implications in a simple palladium-hydrogen system. The diffusion of hydrogen in response to temperature gradients challenges traditional notions of diffusion as it introduces the coupling of thermal and concentration gradients. The resulting driving forces and fluxes are intricately linked, making thermodiffusion a complex but fascinating area of study. Several methods for testing the main parameter used in thermodiffusion, the heat of transport, are implemented using molecular dynamics, and although they don't give similar results, some conditions for isolating the phenomenon have been identified, mainly the importance of controlling the concentration gradient across the material, and the careful control of local volume and pressure.

The aim of the present study is to quantify the impact that local scale physics of phase change has on the overall heat transfer of the quasi-cubic modular element of an innovative Decay Heat Removal System (DHRS). To evaluate the capabilities of using a solid/liquid phase change material (PCM) to drive passive natural circulation to cool the core of a Sodium-cooled Fast Reactor (SFR), a CFD analysis was carried out using STAR-CCM+. The melting-solidification model that utilizes the Volume- Of-Fluid (VOF) enthalpy-porosity approach proved sufficient when liquid and solid phases are isotropic, homogeneous and remain in thermal equilibrium at the interface, however, demonstrated extreme sensitivity when considering variable density between the phases. A preliminary estimation for the placement of the thermocouples according to the calculations was made for a future experimental study of melting Zamak PCM to measure areas of highly convective motion, hot and cold spots, and the shape of the melting interface

Zirconium alloys are the most widely used material for nuclear fuel cladding in light wa- ter reactors where irradiation causes solute redistribution and creation of point defects like vacancies, which degradates the alloy properties. It is mandatory to understand at atomic-scale the transport behavior of the alloying elements in Zr, like Fe. Fe diffusion in Zr is considered anormal because it is much faster than other elements and it does not follow an Arrhenius equation. Previous studies tried to explain this phenomenon relying on the most common and well-knon diffusion mechanisms, where an Fe atom can occupy either an interstitial or substitutional possition, however, they did not man- age to reproduce the experimental results. In this work, a new diffusion mechanism that consists of an Fe-interstitial atom and a vacancy next to it is proposed. DFT cal- culations were performed to calculate binding and migration energies between different configurations of a Fe-vacancy cluster. A semi-analytical code developed at SRMP lab- oratory, KineCluE, was used to calculate the diffusion coefficients of the solute and Zr self-diffusion as well. The results are in accordance to experimental results

Este proyecto surge de un contrato otorgado a IDOM. El objetivo principal de este proyecto es el de evaluar la fiabilidad y disponibilidad del Tritium Extraction System (TES) que forma parte de los sistemas auxiliares de la Test Blanket Module (TBM) de ITER. La introducción de esta tesis proporcionará una visión general de los conceptos y temas funda- mentales que se discutirán. Incluye explicaciones sobre física nuclear, la física de la producción de tritio y detalles técnicos relacionados con los sistemas de ITER. Esto asegura que el lector adquiera una comprensión sólida de los principios básicos antes de adentrarse en discusiones más complejas. Después de las explicaciones introductorias, la tesis profundizará en el concepto de Análisis Probabilístico de Seguridad (PSA, por sus siglas en inglés) y su aplicación única en la industria de la fusión, que difiere del enfoque utilizado en la fisión debido a los riesgos son significativa- mente más bajos, como la radiactividad o el calor residual. Sin embargo, las consideraciones de seguridad siguen siendo muy importantes en la fusión. A continuación, se introducirá el concepto de Reliability, Availability, Maintainability, Inspect- ability (RAMI, por sus siglas en inglés). Este es el enfoque adoptado por la organización ITER, para evaluar la disponibilidad de sus sistemas. Dado que ITER es un experimento pionero con muchos sistemas únicos, garantizar su disponibilidad es crucial para poder mantener la seguri- dad de los equipos especializados. La tesis desarrolla la aplicación del análisis de modos de falla, efectos y criticidad (FMECA, por sus siglas en inglés) en conjunto con el Diagrama de Bloques de Fiabilidad (RBD, por sus siglas en inglés) como herramientas dentro del análisis RAMI. Estas herramientas se describirán en detalle, incluyendo cómo se utilizan para evaluar la criticidad y disponibilidad del sistema, y su papel en el análisis APS de fusión. Una vez que se hayan introducido todos los conceptos relevantes, se llevará a cabo el análisis RAMI del Sistema de Extracción de Tritio para el Helium Cooled Pebble Bed (HCPB) de ITER. Este análisis abarcará varios pasos clave, incluyendo el análisis funcional para comprender pro- fundamente las funcionalidades del sistema, además, se llevará a cabo un análisis de modos de falla, efectos y criticidad (FMECA) para identificar posibles modos de fallo, seguido del desar- rollo de un Diagrama de Bloques de Fiabilidad (RBD) para evaluar la criticidad del sistema y obtener una estimación de la disponibilidad del sistema. Este análisis proporcionará información valiosa sobre la disponibilidad del sistema, así como la identificación de los puntos críticos que requieren una atención especial en términos de fiabilidad y mantenimiento

Deep geological repositories are the most widely accepted international solution to dispose of high activity radioactive wastes. This solution requires an exhaustive analysis of the different containment barriers that will have to isolate these radionuclides for thousands of years so that they are not released into the biosphere prematurely. The last barrier is the geological environment and one of the most important properties that this barrier has is its capacity to retain radionuclides through sorption processes. On the other hand, transuranic elements such as americium or curium are one of the radionuclides with the longest life and radiotoxicity. However, studying these elements is neither easy nor cheap, so the sorption of these elements is usually studied using other more common ones that serve as an analog, europium being the most used. The models used today to describe radionuclide sorption are usually complex, which makes them very sensitive and specific to the conditions used in the experiments, as well as, in some cases, requiring large computational capabilities. Throughout this work we have tried to see how far the sorption models can be simplified. To this aim, we have developed PHREQC-based sorption models capable of explaining europium sorption in illite mineral matrices. As a result, three models have been obtained. Two of which allow simulating some experimental cases without the presence of carbonates within pH ranges between 4 and 10, as well as in concentrations less than 10-6M of europium dissolved in the medium. The last model makes possible to reproduce the sorption of europium in the presence of carbonates within a pH range between 6 and 10, as well as in concentrations less than 10-6M of europium dissolved in the medium.

Transient phenomena are of major importance in the safety analysis of new reactor designs. At the moment, while MYRRHA is in the process of being licensed, engineers and scientists at SCK CEN are performing all the necessary safety studies in the design to ensure it complies with the regulations. Part of such studies are the calculations of the dynamic behaviour of the reactor in transient scenarios, which can range from movements of control rods bundles that take place regularly in the operation of a nuclear reactor to accident conditions. Knowing the time-dependent evolution of the reactivity in the core is essential, as it is a parameter that is required for determining the neutron population, power or thermal-hydraulic behaviour of the reactor, but calculating it is not a simple task. Usually, the approach followed is to approximate the time-dependent reactivity by determining a static reactivity instead, which is not dependant on time and can be more easily calculated. This static reactivity is then used as an input in the point kinetics equations, to measure the evolution of the neutron population, and in thermal-hydraulic codes, to study the feedback effects and other safety-related parameters. However, in the last years the general improvements in computational power and the development of dynamic tools in various Monte Carlo neutron transport codes offer the possibility of calculating the reactivity, as well as other parameters of interest, using a fully time-dependent approach. As Monte Carlo method does not introduce major simplifications in the calculations, they allow to reach best-estimate solutions, which makes the Monte Carlo approach an interesting possibility. In this work, the possibility of using time-dependent Monte Carlo simulations for the safety analysis of the MYRRHA reactor is explored. The approach for performing such calculations, together with case studies using various models are presented. An assessment of the main challenges and shortcomings that need to be overcome for the use of the method in the future is also included. At the moment, though, the Monte Carlo method is not yet applicable in safety studies, and static methods are preferred instead. In an effort to provide the safety team with a reliable tool to calculate the reactivity of the core in a transient situation, the method currently used at SCK CEN is evaluated and compared with the dynamic Monte Carlo simulations. Additionally, an alternative approach for the calculation of the static reactivity, derived from perturbation theory, is proposed and studied

Description: inside the framework of the project for the design and supply of nine leak detection systems for the Vacuum vessel, the Neutral Beam and the cryostat of the ITER facility, the following tasks will be performed: - Revision of the nuclear classification of the nine systems, based on the operating and process data (P&ID's, functional descriptions, operational modes, etc). - Development of the nuclear classification at component level of all the systems, paying special attention to pressure line of the installation. - Monitoring through technical controls of all safety related activities for design, purchasing, manufacturing and assembly.