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Impurity profiles have been measured with the edge high field side (HFS) and low field side (LFS) charge exchange recombination spectroscopy suite at ASDEX Upgrade enabling the study of the poloidal structure of the edge parallel flow in H-mode, L-mode and I-mode. In H-mode, asymmetries in the impurity density, toroidal and poloidal rotations are found. In I-mode, only toroidal rotation asymmetries have been measured while in L-mode no asymmetries have been observed. The measured parallel flow can be divided in two components, the Pfirsch-Schlüter (PS) flow and the symmetric flow. Two different methods have been followed to determine both contributions to the parallel flow. The first method is based on the calculation of the PS flow at the HFS and LFS from the radial electric field. The second method directly provides the symmetric flow from the flux surface average (FSA) of the parallel flow. In H-mode, the methods provide different results, while in L-mode and I-mode they agree. The differences observed in H-mode between the two methods could be explained by the existence of asymmetries in the impurity density, by the non-negligible particle sources and radial losses, or by the approximations made in the calculation of the FSA of the parallel flow from measurements in two poloidal positions (midplane HFS and LFS) only.

One of our grand challenges towards fusion energy is the achievement of a high-performance plasma core coupled to a boundary solution. The high confinement mode (H-mode) provides such a high-performance fusion core due to the build-up of an edge transport barrier leading to a pedestal. However, it usually features type-I edge localized modes (ELMs) which pose a threat for long-duration plasma operation in future fusion devices as they induce large energy fluences onto the plasma facing components and typically are projected to damage the first wall. For future fusion devices, the integration of a stationary no-ELM regime with a power exhaust solution is indispensable. Several no-ELM and small-ELM regimes have extended their operational space in the past years, with the ultimate goal of providing an alternative core–edge solution to ITER and EU-DEMO. Prominent no-ELM or small-ELM alternatives include the I-mode, QH-mode, EDA H-mode, quasi-continuous exhaust (QCE) and ‘grassy’ ELM regimes, X-point radiator scenarios and negative triangularity L-mode. The state-of-the-art, including access conditions and main signatures, of these alternative regimes is reviewed. Many of these regimes partly match the operational space of ITER and EU-DEMO, however, knowledge gaps remain. Besides compatibility with divertor detachment and a radiative mantle, these include extrapolations to high Q operations, low core collisionality, high Greenwald fractions, impurity transport, amongst others. The knowledge gaps and possible strategies to close these gaps to show their applicability to ITER and EU-DEMO are discussed.

A new concept for velocity space thermal ion loss detection is presented. This diagnostic provides pitch angle resolved measurements that are unfeasible with current diagnostics. It uses the same detection principle as the Fast-Ion Loss Detector with a scintillator as the active component and includes a double slit configuration to measure simultaneously the escaping counter- and co-current ions. Simulations show a gyroradius range between 0.15 and 1.00 cm with a resolution below 0.15 cm (for a gyroradius of 1 cm) and a pitch angle range between 30° and 150° with a resolution below 8° for both counter- and co-current ions. The formation of a sheath in front of the detector and its associated electric field may impact the detection principle. Preliminary simulations with a homogeneous electric field show a decrease in the measurable velocity space range, whereas the gyroradius and pitch angle resolution barely change. The strike map is sensitive to the sheath electric field.

In the search for a clean and sustainable energy source for our society, fusion energy emerges as a promising candidate. The realization of a fusion power plant on Earth faces important technological and physical challenges. This thesis is a multidisciplinary project that addresses the optimization of future fusion devices from a plasma physics and engineering perspective. From the plasma physics perspective, the performance of future fusion reactors depends on the properties of the plasma, the fusion fuel. The main ion properties (in present experimental devices, deuterium) are particularly important as they determine the fusion power, which sets the electricity production. Traditionally, the main ions have been rarely diagnosed and their properties have typically been inferred from minority impurity measurements and theoretical models. In this thesis, a novel diagnostic method has been established that enables the direct experimental measurement of the main ions with a focus on the plasma edge. The plasma edge is a critical region, as it prescribes the boundary conditions for the plasma core performance, while it must enable a heat exhaust solution that limits the power loads to the plasma facing components, keeping the integrity of the fusion reactor. Plasma edge diagnostics are demanding in terms of spatial and temporal resolution, as they need to resolve fast transient events and strong spatial gradients. From an engineering perspective, the efficiency of the power conversion cycle coupled to a fusion reactor has been studied and optimized to maximize the electric power output. Cogeneration schemes as opportunities for boosting the efficiency of future fusion power plants have also been investigated. In the framework of this thesis, a new edge main ion diagnostic based on the Charge Exchange Recombination Spectroscopy technique has been installed and exploited at the ASDEX Upgrade experimental reactor, a full metal wall device, to provide main ion temperature a, La fusión nuclear se presenta como una de las grandes opciones tecnológicas para cubrir de manera limpia y sostenible la demanda energética. Sin embargo, lograr la producción de energía en una planta de potencia basada en la fusión nuclear en la Tierra enfrenta importantes desafíos tecnológicos y físicos. Esta tesis es un proyecto multidisciplinar que aborda la optimización de futuros reactores de fusión desde una perspectiva de ingeniería y física del plasma. Desde la perspectiva de la física del plasma, el rendimiento de los futuros reactores de fusión depende de las propiedades del plasma, el combustible de fusión. Las propiedades de la especie iónica mayoritaria (en los dispositivos experimentales actuales, el deuterio) son particularmente importantes ya que determinan la energía producida por la fusión nuclear, que a su vez establece la producción de electricidad. Tradicionalmente, las propiedades del deuterio no se miden de manera directa en reactores experimentales, sino que son inferidas a partir de las propiedades de otras especies iónicas minoritarias (impurezas) y modelos teóricos. En esta tesis, se ha establecido un nuevo método de diagnóstico que permite la caracterización experimental directa de la temperatura y la velocidad de rotación toroidal del deuterio en el borde del plasma. El borde del plasma es una región crítica, ya que prescribe las condiciones de contorno que determinan el desempeño global del plasma y que deben ser compatibles con estrategias de disipación de calor que garanticen la integridad del reactor de fusión. Los diagnósticos de borde de plasma son exigentes en términos de resolución espacial y temporal, ya que necesitan resolver eventos transitorios rápidos y fuertes gradientes espaciales. Desde la perspectiva de la ingeniería, se ha estudiado y optimizado la eficiencia del ciclo de potencia acoplado a un reactor de fusión, maximizando la producción de energía eléctrica. Igualmente, se han investigado esquemas de cogeneración como

We present a non-parametric inference of impurity transport coefficients by using charge exchange recombination spectroscopy measurements of Ne X, Ne VIII, O VIII, and C VI lines. Due to their close atomic numbers, neon, oxygen and carbon impurity ions are assumed to have the same diffusion coefficient D and convection velocity v. Unlike conventional techniques that modulate or perturb the impurity contents, we employ a quasi-stationary plasma with static impurity profiles. Since the ratio of v to D only describes the equilibrated profile of the sum of all impurity charge states, steady-state measurements can still decouple D and v if different charge states are simultaneously observed. We have formulated a non-parametric analysis framework based on the Bayesian probability theory and conducted transport coefficient measurements for a Type III ELMy H-mode plasma at ASDEX Upgrade. The charge exchange reactions with the background neutrals, which are known to affect the impurity charge state balance, are taken into account by introducing additional free parameters. While D at the pedestal is close to the neoclassical level ( < 1 m s-2), a large diffusion coefficient and a strong outward convection are inferred right inside the pedestal top.

A new method was developed to model the neutral population produced by the gas puff based charge exchange recombination spectroscopy systems at ASDEX Upgrade (AUG). With this method, the edge impurity density on the high field side (HFS) and low field side (LFS) can be obtained without the need to apply a neutral beam injection system. The neutral penetration needed for the calculation of the impurity density is obtained with a new gas puff module implemented in the FIDASIM code. The LFS impurity density profile evaluated with the new gas puff module matches the impurity density calculated with standard beam-based charge exchange diagnostics. Impurity temperature, rotation and density profiles at the HFS and LFS of an AUG H-mode discharge are presented. Edge impurity toroidal and poloidal flows show asymmetric structures. The impurity density asymmetries obtained with the new gas puff module are consistent with the observed flow structure.

This paper studies the effect of 3D magnetic perturbations (MPs) on fast-ion confinement, and its impact on the toroidal rotation velocity profile. Two low collisionality H-mode experiments carried out at the ASDEX Upgrade tokamak have been analysed. The two discharges feature different magnetic field helicity (q95), and differences in the velocity-space and level of fast-ion losses are observed. A new analysis technique has been developed that sheds light on the dependencies between fast-ion losses and toroidal rotation, providing for the first time correlation patterns resolved in radius and velocity space of the lost fast-ions. The correlation intensifies towards the plasma edge and is strongly dependent on the orbit topology of the lost fast-ions. The ASCOT orbit following code has been used to characterize the fast-ion resonant transport and beam driven torques, using the vacuum approach and including plasma response (PR). The change of the toroidal canonical momentum, which serves as figure of merit for resonant fast-ion transport, has been calculated with ASCOT. The beam geometry and q95 are found to have a strong impact on the fast-ion transport and losses. The fast-ion transport induced by the MPs affects the beam driven torques. The effect of the changes of the j × B and collisional torques on plasma rotation is analysed using the torques simulated by ASCOT and simple momentum balance calculations. For the low q95 = 3.8 discharge, which benefits from a resonant amplification, we find excellent agreement with the measured variation of the toroidal velocity. For the high q95 = 5.5 discharge, the inclusion of the PR improves the comparison with experimental data with respect to the vacuum estimation, but still some differences with experiments are observed. This suggests that other non-resonant effects could play a role for the determination of the toroidal rotation profile.

We present a non-parametric inference of impurity transport coefficients by using charge exchange recombination spectroscopy measurements of Ne X, Ne VIII, O VIII, and C VI lines. Due to their close atomic numbers, neon, oxygen and carbon impurity ions are assumed to have the same diffusion coefficient D and convection velocity v. Unlike conventional techniques that modulate or perturb the impurity contents, we employ a quasi-stationary plasma with static impurity profiles. Since the ratio of v to D only describes the equilibrated profile of the sum of all impurity charge states, steady-state measurements can still decouple D and v if different charge states are simultaneously observed. We have formulated a non-parametric analysis framework based on the Bayesian probability theory and conducted transport coefficient measurements for a Type III ELMy H-mode plasma at ASDEX Upgrade. The charge exchange reactions with the background neutrals, which are known to affect the impurity charge state balance, are taken into account by introducing additional free parameters. While D at the pedestal is close to the neoclassical level ( < 1 m s-2), a large diffusion coefficient and a strong outward convection are inferred right inside the pedestal top.

A new method was developed to model the neutral population produced by the gas puff based charge exchange recombination spectroscopy systems at ASDEX Upgrade (AUG). With this method, the edge impurity density on the high field side (HFS) and low field side (LFS) can be obtained without the need to apply a neutral beam injection system. The neutral penetration needed for the calculation of the impurity density is obtained with a new gas puff module implemented in the FIDASIM code. The LFS impurity density profile evaluated with the new gas puff module matches the impurity density calculated with standard beam-based charge exchange diagnostics. Impurity temperature, rotation and density profiles at the HFS and LFS of an AUG H-mode discharge are presented. Edge impurity toroidal and poloidal flows show asymmetric structures. The impurity density asymmetries obtained with the new gas puff module are consistent with the observed flow structure.

Nuclear fusion is expected to be a clean and almost-unlimited power source in the near future. The first net power demonstration plant (DEMO) is planned to start operation in 2050. The supercritical carbon dioxide (S-CO2) Brayton cycle is an excellent candidate for integration with a fusion power plant, such as DEMO, because of its high efficiency at intermediate temperatures and low interaction of coolant with tritium. This work analyses a set of S-CO2 Brayton cycle layouts for its integration in a DEMO-like fusion power plant, considering the specific requirements and heat availability characteristics. A framework has been developed to integrate the PROCESS code and the numerical solver EES to study the thermal and economic aspects of integrating the different S-CO2 cycle layouts. In total, 14 layouts have been studied and grouped into a more conservative (DEMO1, pulsed operation) and more advanced (DEMO2, steady-state operation) fusion reactors. The PROCESS code has been used to obtain the DEMO 2018 Baseline, which defines the available power from each heat source and their boundary conditions. This code has also been used to assess the cost of the optimal layout. Thermal storage has been added to the DEMO1 scenario to avoid standby times that could negatively affect the cycle equipment lifetime and efficiency. Besides, these boundary conditions have been extended to account for possible technical improvements by the time of its construction in the DEMO2 scenario. A sensitivity analysis of the most characteristic parameters of the cycles shows a strong dependence on the turbine inlet temperature for all layouts, which is constrained by the reactor material limits. The cycle efficiency (electric power produced before consumptions non-related to the cycle) has been selected as the figure of merit for the optimisation. The results show a 38% cycle efficiency for DEMO1 and 56% for DEMO2 scenarios. These efficiencies drop to 20% and 38% values, respectively, when the

Nuclear fusion is expected to be a clean and almost-unlimited power source in the near future. The first net power demonstration plant (DEMO) is planned to start operation in 2050. The supercritical carbon dioxide (S-CO) Brayton cycle is an excellent candidate for integration with a fusion power plant, such as DEMO, because of its high efficiency at intermediate temperatures and low interaction of coolant with tritium. This work analyses a set of S-CO Brayton cycle layouts for its integration in a DEMO-like fusion power plant, considering the specific requirements and heat availability characteristics. A framework has been developed to integrate the PROCESS code and the numerical solver EES to study the thermal and economic aspects of integrating the different S-CO cycle layouts. In total, 14 layouts have been studied and grouped into a more conservative (DEMO1, pulsed operation) and more advanced (DEMO2, steady-state operation) fusion reactors. The PROCESS code has been used to obtain the DEMO 2018 Baseline, which defines the available power from each heat source and their boundary conditions. This code has also been used to assess the cost of the optimal layout. Thermal storage has been added to the DEMO1 scenario to avoid standby times that could negatively affect the cycle equipment lifetime and efficiency. Besides, these boundary conditions have been extended to account for possible technical improvements by the time of its construction in the DEMO2 scenario. A sensitivity analysis of the most characteristic parameters of the cycles shows a strong dependence on the turbine inlet temperature for all layouts, which is constrained by the reactor material limits. The cycle efficiency (electric power produced before consumptions non-related to the cycle) has been selected as the figure of merit for the optimisation. The results show a 38% cycle efficiency for DEMO1 and 56% for DEMO2 scenarios. These efficiencies drop to 20% and 38% values, respectively, when the rea

The dynamic behaviour of the ion and electron energy, particle and momentum transport measured during type-I edge localized mode (ELM) cycles at ASDEX Upgrade is presented. Fast measurements of the ion and electron temperature profiles revelead that the ion and electron energy transport recover on different timescales, with the electrons recovering on a slower timescale (Cavedon et al 2017 Plasma Phys. Control. Fusion 59 105007). The dominant mechanism for the additional energy transport in the electron channel that could cause the delay in the electron temperature gradient (VTe) recovery is attributed to the depletion of energy caused by the ELM. The local sources and sinks for the electron channel in the steep gradient region are much smaller compared to the energy flux arriving from the pedestal top, indicating that the core plasma may dictate the local dynamics of the VTe recovery during the ELM cycle. A model for the edge momentum transport based on toroidal torque balance that takes into account the existence of poloidal impurity asymmetries has been developed. The analysis of the profile evolution during the ELM cycle shows that the model captures the dynamics of the rotation both before the ELM crash and during the recovery phase.

Within the EUROfusion MST1 Work Package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (ne,sep/nG ∼ 0.3), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a Double Null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened.

The upgrade of the high field side (HFS) edge charge exchange recombination spectroscopy (CXRS) system of ASDEX Upgrade is presented. This diagnostic provides temperature, rotation and radiance measurements of impurity species by taking advantage of the gas puff based CXRS technique (GP-CXRS). The system is formed by a fast piezoelectric valve, that injects thermal neutrals into the plasma, and two optical heads. The localized gas injection together with properly aligned lines of sights (LOS) lead to a high spatial resolution of 5–19 mm. Fast gas puff modulation allows a precise subtraction of the passive part of the signal. The existing poloidal optical head has been replaced with a new one to increase the radial resolution. The number of lines of sight (LOS) of the poloidal optical head has been increased from 8 to 16 covering around 7 cm of the plasma edge at the HFS. The same radial range is also viewed by a toroidal optical head. The neutral deposition, needed to calculate the impurity density profile, has been modelled using the FIDASIM code. A realistic gas puff geometry has been implemented in the code. The first measurements of impurity temperature, rotation and radiance utilizing the upgraded diagnostic are presented.

The edge ion heat transport is analyzed in ASDEX Upgrade (AUG) by combining a comprehensive set of pedestal measurements with both interpretive and predictive modelling. The experimentally determined ion heat diffusivities, $\chi_{\rm i}$ , are compared with neoclassical theory and the impact of edge localized modes (ELMs) on the edge ion heat transport level is studied in detail. Pedestal matching experiments in deuterium and hydrogen plasmas show that the inter-ELM pedestal $\chi_{\rm i}$ remains close to the neoclassical value. The additional power needed in hydrogen to get similar pedestal temperatures as in deuterium plasmas mostly affects the electron heat channel, i.e. the electron heat diffusivity increases while the ion heat diffusivity stays at the same level within the uncertainties. Sub-ms measurements of the edge ion temperature allows us to extend the analysis to the entire ELM cycle. During the ELM crash, the ion heat transport is increased by an order of magnitude. The perturbed heat flux increases first at the separatrix, i.e. first the separatrix ion temperature increases, leading to a flatter ion temperature gradient, followed by a decrease of the whole pedestal profile. The ion heat transport returns to its pre-ELM neoclassical level 3–4 ms after the ELM crash.

El modo de alto confinamiento (H-mode) está considerado como el modo de operación óptimo para futuros dispositivos de fusión nuclear. El H-mode viene acompañado de inestabilidades que provocan la expulsión de partículas y energía, y deterioran el confinamiento del plasma. Aunque dichas inestabilidades son tolerables en los dispositivos actuales, el desarrollo de t´ecnicas para su control es fundamental de cara a garantizar la integridad de futuros reactores. La aplicación de perturbaciones magnéticas externas es una de las herramientas más extendidas para la mitigación e incluso completa supresión de dichas inestabilidades. Estas perturbaciones alteran la simetría del campo magnético que confina al plasma. La respuesta del plasma a dichas perturbaciones es una línea de investigación muy activa actualmente. El objetivo de este trabajo es el estudio detallado de dos descargas de baja colisionalidad, llevadas a cabo en el tokamak ASDEX Upgrade. Durante estas descargas se aplican perturbaciones magnéticas externas para estudiar la respuesta del plasma. El estudio está centrado en el análisis de la evolución temporal acoplada de la rotación toroidal de las impurezas, pérdida de iones rápidos y densidad. Los resultados obtenidos revelan fuertes correlaciones entre estos para´metros. Las interdependencias entre las magnitudes dependen de (1) la caracterización de la población de iones rápidos sujeta a pérdidas y (2) la posición radial en la que se miden rotación y densidad. Se han observado fuertes cambios en el patrón de correlaciones asociados a las diferencias en las órbitas de los iones rápidos; y la intensificación de la correlación hacia el borde del plasma. Bajo determinadas circunstancias, se produce un cambio en el sentido de giro de la rotación toroidal de las impurezas en el borde del plasma, i.e. las impurezas en el borde y en el centro giran en sentido contrario. Ello también conlleva un cambio en el patrón de la correlación., The high-confinement mode (H-mode) is considered to be the baseline operational scenario for future nuclear fusion devices. The H-mode is accompanied by instabilities that cause cyclic ejection of particles and energy, thus deteriorating the confinement of the plasma. Although these instabilities are tolerable in current experiments, the development of techniques for their control is crucial to guarantee the integrity of future devices. The application of external 3D magnetic fields is one of the most efficient techniques to achieve the mitigation and even full suppression of these instabilities. These magnetic perturbations break the symmetry of the magnetic field. The impact of the perturbations on the plasma is currently a very active research line. The objective of this thesis is a detailed analysis of two low collisionality discharges, which were carried out in the ASDEX Upgrade tokamak experiment. In these two discharges static magnetic perturbations were applied to study plasma response. The study is focused on the observation of the coupled evolution of the toroidal impurity rotation, fast-ion losses and electron density. The results reveal strong correlations between the parameters. The interdependencies between the parameters depend on (1) the type of fast-ion population which is subject to losses and (2) the radial position at which rotation and density are measured. A strong change in the correlation pattern is associated with changes in the orbit topology of the fast-ions. The correlation towards the plasma edge becomes more intense. Under certain circumstances, a change in the correlation pattern at the edge is connected to an impurity toroidal rotation reversal, i.e. a spin-up of the impurities in opposite direction to the plasma core.

[ES] En el Trabajo de Fin de Grado, se ha desarrollado un modelo reducido de un reactor nuclear de agua a presión para la simulación de transitorios de inserción de reactividad, tanto positiva como negativa, que es uno de los tipos de accidentes incluidos en las bases de diseño de esta tecnología. El análisis del comportamiento de la planta utilizando modelos reducidos permite conocer, de manera general, la evolución de la planta de manera rápida. Se ha construido el modelo matemático considerando los fenómenos físicos que dominan la neutrónica, haciendo uso de las ecuaciones de la cinética puntual, y termohidráulica del reactor. El modelo obtenido es un sistema de ecuaciones diferenciales, que dan cuenta de la evolución de las variables más representativas en términos de seguridad de operación de la planta nuclear. Destaca la importancia del seguimiento de la potencia térmica y la temperatura de combustible, como variables más críticas desde el punto de vista de la seguridad de la planta. El modelo ha sido implementado en Matlab, dados su extendido uso científico y su conveniencia para la resolución del sistema de ecuaciones diferenciales planteado. Tras la ejecución del código en estado estacionario y su verificación, se ha procedido a la simulación de los transitorios. Los resultados de las evoluciones de las variables más importantes obtenidos con el modelo reducido han sido comparados con datos de referencia, extraídos mediante el código detallado RELAP5, avalado por organismos reguladores en el ámbito nuclear, para comprobar su adecuación. De esta comparación se concluye que el modelo reducido desarrollado es apropiado, puesto que proporciona unos buenos resultados comparados con la referencia., [CA] En el Treball de Fi de Grau, s’ha desenvolupat un model reduït d’un reactor nuclear d’aigua a pressió per a la simulació de transitoris d’inserció de reactivitat, tan positiva com negativa, que és un dels tipus d’accidents inclosos en les bases de disseny d’aquesta tecnologia. L’anàlisi del comportament de la planta utilitzant models reduïts permet conèixer, de manera general, l’evolució de la planta de manera ràpida. S’ha construït el model matemàtic considerant els fenòmens físics que dominen la neutrònica, fent ús de les equacions de la cinètica puntual, i termo-hidràulica del reactor. El model obtingut és un sistema d’equacions diferencials, que donen compte de l’evolució de les variables més representatives en termes de seguritat d’operació de la planta nuclear. Destaca la importància del seguiment de la potència tèrmica i la temperatura de combustible, com variables més crítiques des del punt de vista de la seguritat de la planta. El model ha sigut implementat en Matlab, donat el seu estès ús científic i la seua conveniència per a la resolució del sistema d’equacions diferencials plantejat. Després de l’execució del codi en estat estacionari i la seua verificació, es va procedir a la simulació dels transitoris. Els resultats de les evolucions de les variables més importants obtinguts amb el model reduït han sigut comparats amb dates de referència, extrets mitjançant el codi detallat RELAP5, avalat per organismes reguladors en l’àmbit nuclear, per comprovar la seua adequació. D’aquesta comparació es conclou que el model reduït desenvolupat és adequat ja que proporciona uns bons resultats comparats amb la referència., [EN] In this Bachelor Thesis, a reduced model of a nuclear pressurized water reactor for simulation of reactivity insertion transients, both positive and negative, has been developed. This type of transient is included in the accident design basis of the given technology. The analysis of plant’s performance using reduced models allows to evaluate the general evolution of the plant in a quick manner. A mathematical model which considers point kinetic equations for power calculation and thermo-hydraulics processes in different parts of the reactor has been constructed. The model consists of a differential equations system, which account for the evolution of the most representative variables in terms of nuclear safety. It is remarkably important the monitoring of thermal power output and fuel temperature as most critical variables regarding nuclear safety in these transients. The model has been implemented in Matlab, given its widespread scientific use and suitability for solving the raised differential equations system. After the steady state code execution and verification, transients’ simulations have been carried out. The results of the evolution of the most important variables obtained with the reduced model have been compared with reference data extracted from RELAP5 detailed code, which is a code supported by regulatory bodies in the nuclear field. From this comparison it is concluded that the reduced model developed is appropriate to analyze reactivity insertion transients as the obtained results agree with the ones provided by the reference code.