System codes and their necessary power plant nodalizations are an essential step in thermal hydraulic safety analysis. In order to assess the safety of a particular power plant, in addition to the validation and verification of the code, the nodalization of the system needs to be qualified. Since most existing experimental data come from scaled-down facilities, any qualification process must therefore address scale considerations. The Group of Thermal Hydraulic Studies at Technical University of Catalonia has developed a scaling-up methodology (SCUP) for the qualification of full-scale nodalizations through a systematic procedure based on the extrapolation of post-test simulations of Integral Test Facility experiments. In the present work, the SCUP methodology will be employed to qualify the nodalization of the AscóNPP, a Pressurized Water Reactor (PWR), for the reproduction of an important safety phenomenon which is the effectiveness of the Core Exit Temperature (CET) as an Accident Management (AM) indicator. Given the difficulties in placing measurements in the core region, CET measurements are used as a criterion for the initiation of safety operational procedures during accidental conditions in PWR. However, the CET response has some limitation in detecting inadequate core cooling simply because the measurement is not taken in the position where the cladding exposure occurs. In order to apply the SCUP methodology, the OECD/NEA ROSA-2 Test 3, an SBLOCA in the hot leg, has been selected as a starting point. This experiment was conducted at the Large Scale Test Facility (LSTF), a facility operated by the Japanese Atomic Energy Agency (JAEA) and was focused on the assessment of the effectiveness of AM actions triggered by CET measurements. The steps of the SCUP methodology are presented: post-test calculation, scaling effect analysis and design effect analysis. The AscóNPP nodalization has shown to be qualified for the simulation of the involved phenomenology. The final step of the work presented here was to adapt the boundary conditions to a more realistic situation taking place in the AscóNPP. CET and Peak Cladding Temperature (PCT) readings were seen to present large differences similarly as it occurred in the ROSA-2 Test 3. The PCT when all the CET readings were credited was 872 K, however, when only the minimum CET reading was credited the maximum temperature in the core rose to 1053 K.
Martinez, V.; Freixa, J.; Perez, M.; Reventós, F. International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety p. 1-12 Data de presentació: 2016-10 Presentació treball a congrés
In the framework of Design Basis Accidents (DBA) for Pressurized Water Reactor (PWR), and based on recent studies on pipe integrity combined with the Risk-Informed Decision Making (RIDM), the USNRC proposed the Intermediate Break LOCA (IBLOCA) scenario to become a design basis accident. These studies reported that the probability of a complete rupture of a pipe depends on the pipe size, and in particular that this probability is higher for smaller tubes. Therefore, the double guillotine break of smaller pipes connected to the primary side of a PWR such as the surge line or the ECCS lines should pose a more probable scenario than an integral rupture of the cold or hot legs. In order to assess the effectiveness of the safety strategies and the Emergency Operational Procedures (EOP) for a selected Nuclear Power Plant (NPP) and scenario, Best Estimates (BE) codes like RELAP5 are of great value because they allow simulating the overall behavior and response of the system under accident conditions. Furthermore, for licensing purposes, BE simulations should be combined with uncertainty analyses to yield the so-called Best Estimate Plus Uncertainty (BEPU) calculations. In the present paper a preliminary assessment of an IBLOCA BEPU calculation is carried out for the Asc ó -2 NPP (3-loop PWR Westinghouse design). The transient follows its reported EOPs with Emergency Core Cooling Systems (ECCS) failure assumptions for core uncover conditions. A preliminary evaluation of the effectiveness in the selection of the input uncertainty parameters is presented. Those that were concluded as relevant for LBLOCA and SBLOCA in previous works are analyzed to determine which are the most influential to be considered in IBLOCA. In addition, the influence of the nodalization qualification is also studied. Two different nodalizations are compared for assessing the significance of the modelling approach in BEPU analyses: the first one, a 1D vessel nodalization that was qualified with operational events reported in the actual NPP; and the second one, the same nodalization with a Pseudo 3D modelling of the vessel (core parallel channels with transversal flow paths and DC parallel channels with azimuthal connections). This second nodalization was qualified for SBLOCA accidents with experiments performed at Integral Test Facilities (G7.1 of PKL and Test 3 of LSTF ) b y the use scaling techniques (SCUP methodology).
Freixa, J.; Martinez, V.; Reventós, F. International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety p. 1/14-14/14 Data de presentació: 2016-10 Presentació treball a congrés
After the Fukushima accident, “st ress- test” activities carried out worldwide pointed out the need to study additional accident management measures to deal with prolonged Station Black Out (SBO) scenarios. Without any operator actions, a total loss of the secondary side heat sink leads to core uncovery, to core damage and ultimately to a melt -down scenario. The international NEA/OECD PKL -3 project has addressed the efficiency of possible accident management actions to re- establish core cooling by experiments at the PKL and PMK test faciliti es. The experiments at the PKL test facility were focused on the efficiency of the counter measures to re-establish the core cooling and the performance of core exit temperature measurements during the core heat up phase. In this type of scenarios, the cor e power has been significantly reduced; there are no break locations and only either sporadic or temporary openings of relief or safety valves. In this situation, sources of momentum are much smaller than in typical LOCA or SBLOCA events. Since best estimate system codes were mainly developed to simulate LOCA scenarios (in their full spectrum) the performance of system codes and the general guidelines followed to simulate PWR power plants are called into question. In this paper, RELAP5 simulations of two SB O experiments of the PKL facility are presented. Firstly, a post -test analysis of Test H2 -1 is presented and specific guidelines on modelling are given. Secondly, these guidelines are applied to the PKL nodalization to blindly simulate Test H2 -2. Finally, a combined post -test for both experiments has been performed. The RELAP5 system code has been assessed for its use in this type of scenarios.
Experimental results obtained at integral test facilities are used in the validation process of thermal - hydraulic system codes for the steady - state and transient simulation of light water reactors. The expertise and guidelines derived fro m this work (nota bly the nodalis ation scheme and the specific model options) can later be applied to safety analyses of nuclear power plants. This paper describes work carried out within the STARS project at the Paul Scherrer Institut (PSI) using the U.S. NRC - sponsored sys tem code TRACE. Post - test analyses for eight selected ROSA and ROSA - 2 experiments from the OECD/NEA projects ROSA and ROSA - 2 were conducted previ ously using TRACE version 5.0 release candidate (RC ) 3 . A consistent plant nodalisation is employed across all tests, which include typical design and beyond design basis accidents, such as loss - of - coolant - accident, main steam line break, and steam generator tube rupture. The full series of cases has been consolidated and upgraded for TRACE version 5.0 Patch 4. Sin ce the newer TRACE version includes substantial changes in the area of choked flow modelling, the nodalisation of the break valves has proven to be a critical upgrade in all test cases. Selected results for the newer TRACE versi on are presented and discuss ed. The complete set of tests has been supplemented by one further post - test analysis presented herein; The OECD/NEA ROSA - 2 Test 7, a PWR 13% cold leg intermediate break los s - of - coolant accident (IBLOCA). Results show that, for this break size , the influen ce of the uncertainty in the break flow rate is substantial. T here is a fine balance between the timing of the core uncovery and loop seal clearance, the rate of coolant loss following core uncovery, and the ability of the loop seal clearance to effectivel y quench the core. This has a significant effect on the predicted peak cladding temperatures. Updated results for ROSA - 2 a nalysis results for the remaining test cases are consistent with the previous RC3 analyses and match the experimental data well . U pdat ed results for ROSA - 2 Test 4, a steam generator tube rupture scenario, show improved agreement with experimental data . KEYWORDS TRACE, ROSA/LSTF, ITF, V&V
A raíz del accidente de la central nuclear de Fukushima Daiichi, se han realizado pruebas de resistencia encaminadas al estudio de las medidas y acciones a llevar a cabo durante escenarios de pérdida de suministro eléctrico exterior prolongado (SBO) sin acción de los operadores. En las centrales nucleares del tipo PWR, la pérdida total de la potencia de refrigeración del sistema secundario conlleva, en ausencia de medidas de gestión de accidentes, un descubrimiento del núcleo, que finalmente desembocará en la fusión parcial o total del combustible. El proyecto internacional PKL-3 de la Organización para la Cooperación y el Desarrollo Económico (OCDE) y la Agencia de la Energía Nuclear (NEA) ha abordado, mediante experimentos en la instalación experimental PKL, la efectividad de las posibles acciones para mantener la refrigeración del núcleo en caso de un eventual SBO. En este tipo de escenarios, la potencia se ha reducido sustancialmente, no hay grandes roturas, solo aperturas de válvulas del presionador. Por lo tanto, las fuentes de cantidad de movimiento en el sistema primario son mucho menores que para los típicos escenarios LOCA. Dado que los códigos de sistema fueron desarrollados básicamente para la simulación de casos LOCA en todo su espectro, es necesaria una validación de los códigos para su uso en casos de SBO. En este trabajo se presentan simulaciones de dos experimentos SBO del proyecto PKL-3 realizadas con el código termo hidráulico RELAP5. Los resultados obtenidos representan correctamente la evolución de los principales parámetros de los experimentos. En el artículo, se destacan las líneas generales que se han seguido para reproducir correctamente la fenomenología de los experimentos.
The increasing importance of Best-Estimate Plus Uncertainty (BEPU) analyses in nuclear safety and licensing processes have lead to several international activities. The latest findings highlighted the uncertainties of physical models as one of the most controversial aspects of BEPU. This type of uncertainties is an important contributor to the total uncertainty of NPP BE calculations. Due to the complexity of estimating this uncertainty, it is often assessed solely by engineering judgment. The present study comprises a comparison of two different state-of-the-art methodologies CIRCÉ and IPREM (FFTBM) capable of quantifying the uncertainty of physical models. Similarities and differences of their results are discussed through the observation of probability distribution functions and envelope calculations. In particular, the analyzed scenario is core reflood. Experimental data from the FEBA and PERICLES test facilities is employed while the thermal hydraulic simulations are carried out with RELAP5/mod3.3. This work is undertaken under the framework of PREMIUM (Post-BEMUSE Reflood Model Input Uncertainty Methods) benchmark.
The recent accident in the Fukushima Daiichi nuclear power plant opened a discussion on severe accident management that includes the analysis of the accident by means of computational tools that can predict the core behavior in such extreme conditions. The RELAP/SCDAPSIM/MOD3.5 code is designed to predict the behavior of Light Water Reactor (LWR) coolant systems during normal and accident conditions including severe accidents up to the point of reactor vessel failure. The code consists of two parts: the RELAP5 models calculate the overall Reactor Coolant System (RCS) thermal-hydraulic response, control system behavior, reactor kinetics and the behavior of special reactor system components such as valves and pumps, to predict the plant behavior under operational transients, Design Basis Accidents (DBAs) and Beyond DBAs; the SCDAP models calculate the behavior of the core and vessel structures under normal and severe accident conditions. Both portions of the code have been proven, separately, to accurately reproduce the response under its designed purpose, which are steady state, DBAs and BDBAs for the RELAP portion, and steady state and severe accident conditions for the SCDAP portion. The analysis of potential scenarios does not define a priori the final state of the fuel rods, and thus the most adequate tool is a system code such as RELAP/SCDAPSIM/MOD3.5 capable of simulating accident scenarios where severe accident phenomena may or may not occur. The present paper revisits the ISP-13 exercise, a cold leg double-ended guillotine LOCA conducted in the LOFT experimental facility, using two RELAP/SCDAPSIM/MOD3.5 models: the first one is entirely modeled with RELAP components, the second model keeps the RELAP nodalization with the exception of the core region, which is modeled with SCDAP components. The LOFT L2.5 experiment is a rather unique experiment since it features nuclear (UO2) fuel rods in a facility designed to simulate the major responses of a commercial pressurized water reactor (PWR). In addition, the fuel cladding of this experiment reached relatively high temperatures of around 1100 K. Even though this cladding temperature is far from the oxidation onset with steam, the LOFT L2-5 experiment challenges system behavior simulations by bringing the conditions close to those of severe accidents. The final goal is to evaluate whether the use of SCDAP components in LOFT L2-5 experiment reproduces similar results to those obtained with a RELAP standalone model, and that both simulations are in good agreement with experimental data.
The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. Thermal hydraulic safety analyses are being carried out with the system code RELAP/SCDAPSIM/MOD4.0 whic h was initially designed to predict the behaviour of light water reactor systems during normal and accident conditions. The code is being developed as part of the international SCDAP Development and Training Pr ogram (SDTP) coordinated by Innovative Systems Software (ISS). The modeling strategy of the RELAP code for the simulation of two-phase flows is based on a single-fluid two-phase approach with a set of momentum, energy and mass equations for each phase. The two phases are liquid-water and gas phase mixture of steam and non-condensable gases. Phase interactions, such as interphase friction and heat transfer, are modelled by closure relations based on experimental/numerical correlations that depend on the flow regime. In cooperation with ISS, the IPR team has implemented LLE liquid phase thermodynamic properties as a working fluid alternative to water and appropriate wall-to-LLE heat transfer correlations. However, in order to analyze some of the postulated off-normal events, there is a need to simulate the mixing of helium and Lead-Lithium fluids In the standard RELAP/SCDAPSIM/MOD4.0 ve rsion it is not possible to simulate a mixture of a non-water fluid with a non-condensable. In addition to that, t he available flow regime maps for vertical and horizontal flows in RELAP are specific for steam/water pair, which may not be suitable for LLE/helium pair. The Technical University of Catalonia is cooperating with IPR to adapt the RELAP/SCDAPSIM/MOD4.0 code to allow the si mulation of LLE and he mixture. This paper presents the results of the first step of the project, which includes a state of the art on simulation of liquid metals mixed with non-condensable using system codes, the implementation of the necessary code modifications to allow for a LLE/he mixture a nd preliminary results using the modified code version for horizontal and vertical configurations.
Given the difficulties in placing measurements in the core region, core exit temperature (CET) measurements are used as a criterion for the initiation of safety operational procedures during accidental conditions in pressurized water reactors (PWR). However, the CET response has some limitation in detecting inadequate core cooling simply because the measurement is not taken in the position where the cladding exposure occurs. The Group of Thermal Hydraulic Studies of the Technical University of Catalonia has conducted analytical studies to assess the performance of RELAP5 and the nodalization approaches for CET predictions through post-test analyses of the OECD/NEA ROSA-2 Test 3 experiment. This experiment was conducted at the Large Scale Test Facility (LSTF), a facility operated by the Japanese Atomic Energy Agency (JAEA), and represented an SBLOCA in the hot leg. The studies carried out have led to deriving an updated nodalization approach for the core region and upper plenum. The knowledge acquired with post-test analyses has been transferred to a full plant model of the Ascó nuclear power plant (NPP) through Kv scaling calculations. The scalability between the LSTF and the Ascó NPP has been analyzed. The necessary changes in the nodalization to correctly reproduce the CET response, as indicated by the post-test calculations, have been added to the Ascó NPP model. The final step of the work presented here was to adapt the boundary conditions to a more realistic situation taking place in the Ascó NPP. CET and PCT readings were seen to present large differences similarly as it occurred in the ROSA-2 Test 3. When the CET reached the safety criteria, the PCT measured was in the range of [777, 906] K depending on which CET measurement was considered as a reference. Due to the high temperatures at the time the set point is triggered, the effectiveness of the accident management actions are at stake and therefore future studies should be focused on the analysis of the evolution of the scenario after the CET signal is reached and the assessment of the CET set-point value.
Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT).; The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects.; The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.
Batet, L.; Mas de les Valls, E.; Osychenko, O.; Martinez, V.; Perez, M.; Freixa, J.; Reventos, F. Reunión Anual de la Sociedad Nuclear Española p. 1-7 Data de presentació: 2014-10-02 Presentació treball a congrés
El Grupo de Estudios Termohidráulicos (GET) de la UPC posee gran experiencia en estudios de seguridad nuclear mediante códigos de simulación de plantas nucleares. Desde 1987, el grupo ha colaborado con las CN de Ascó y Vandellòs II en la cualificación de modelos integrales de planta y dando apoyo a actividades de operación y control. En los últimos años (desde 2006), GET está involucrado en investigaciones relacionadas con la termohidráulica de las centrales nucleares de fusión, ámbito en el cual adopta la etiqueta T4F.
GET/T4F está adaptando y desarrollando herramientas predictivas en diferentes escalas fenomenológicas que, combinadas, permitirán el análisis de fenómenos complejos, como el transporte de tritio por el sistema completo del reactor, la optimización de las estrategias de control de potencia (incluyendo puesta en marcha y seguimiento de carga), la caracterización de las cargas térmicas en diferentes materiales estructurales y la simulación de escenarios accidentales hipotéticos, entre otras.
Las actividades de T4F pueden agruparse en tres grandes áreas:
Descripción microscópica, mediante técnicas de Montecarlo y Dinámica Molecular, de la nucleación de helio en el interior de los canales de metal líquido de una envoltura regeneradora de tritio.
Análisis del comportamiento termohidráulico de una envoltura regeneradora con metal líquido. Se ha desarrollado el código CFD de fuente libre OpenFOAM para aplicarlo al estudio de: acoplamiento MHD-térmico; inestabilidades de flujo; transporte de tritio; transporte y nucleación de helio en el metal líquido; efecto de la presencia de burbujas de helio sobre la transferencia de calor, caída de presión y permeación de tritio.
Análisis integral mediante el código termohidráulico de sistema RELAP5-3D/ATHENA, desarrollado por el Idaho National Laboratory. El código se ha aplicado a la simulación de la planta, incluyendo el sistema de conversión de energía térmica a eléctrica (un ciclo de CO2 supercrítico) y se han analizado diferentes transitorios.
The performance of the RELAP5 thermal-hydraulic system code was analyzed in predicting very fast transient condensation processes in horizontal pipes. The code significantly underpredicted the heat transfer from the primary to the secondary side in case of rapid wall condensation process in the so called Inverse Edwards Pipe Experiment, where the condensation pipe was immerged in a cool water pool, and hot steam injection was performed into a pipe, which was closed on one side. The RELAP5 condensation model for horizontal pipes was modified in order to take into account a stratified flow pattern, and the effect of the local void fraction. The modified RELAP5 code was compared to the original code through the calculation of the Inverse Edwards Pipe Experiment. An improved prediction of the heat transfer process was achieved, considering the temperature, pressure and void fraction distribution along the horizontal pipe during the transient condensation process. (C) 2014 Elsevier B.V. All rights reserved.
In the framework of the nodalization qualification process and quality guarantee procedures and following the guidelines of Kv-scaled analysis and UMAE methodology, further development has been performed by UPC team resulting in a scaling-up methodology. Such methodology has been applied in this paper for analyzing discrepancies that appear between the simulations of two counterpart tests. It allows the analysis of scaling-down criterion used for the design of an ITF and also the investigation of the differences of configuration between an ITF and a particular NPP. For analyzing both, it applies two concepts
The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility (LSTF). Two intermediate break loss-of-coolant-accident (LOCA) experiments (Tests 1 and 2) have been carried out during 2010. The two tests were analyzed by using the US-NRC TRACE best estimate code. A post-test calculation was performed for each test along with uncertainty analysis providing uncertainty bands for each relevant time trend. Uncertainties in the code modelling capabilities as well as in the initial and boundary conditions were taken into account, following the guidelines and lessons learnt through participation in the OECD/NEA BEMUSE programme. Two different versions of the TRACE code were used in the analysis, providing a qualitatively good prediction of the tests.
The OECD/NEA ROSA-2 project aims at addressing thermal-hydraulic safety issues relevant for light water reactors by building up an experimental database at the ROSA Large Scale Test Facility. Test 1 of the ROSA-2 program consists of an intermediate break located in one of the hot legs, in particular it represents the rupture of the pressurizer surge line. Using a TRACE model, a blind case calculation was performed. An uncertainty analysis was carried out together with the simulation of Test 1, in order to provide uncertainty bands for each time trend and finally determine whether the TRACE simulation is able to capture the experimental results within the uncertainty bands. Since the uncertainty bands did not envelop the experimental data, a post-test analysis was carried out. The post-test analysis was helpful in determining which relevant physical phenomena had not been included in the pre-test analysis. The new uncertainty bands derived after including all relevant phenomena enveloped the experimental data.
The pressurized water reactor APR 1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations.
Experimental results obtained at integral test facilities (ITF) are used in the validation process of system codes for the transient analyses of light water reactors. The expertise and guidelines derived from this work are later applied to transient analyses of nuclear power plants (NPPs). However, the boundary conditions at the NPPs will always differ from those at the ITF, and hence, the soundness of the ITF model needs to be maximized. An unaltered ITF nodalization should prove to be able to simulate as many tests as possible, before any conclusion is derived to NPP analyses. Tests carried out at the LSTF facility operated by the Japan Atomic Energy Agency have been simulated in recent years by using the US-NRC system code TRACE. In this paper, 5 different post-test analyses are presented, along with the evolution of the employed TRACE nodalization and the process followed to track the consistency of the modifications.
In cooperation with the Finnish Radiation and Nuclear Safety Authority (STUK), a project has been launched at the Paul Scherrer Institute (PSI) aimed at performing safety evaluations of the Olkiluoto-3 nuclear power plant (NPP), the first EPR™, a generation III pressurizer water reactor (PWR); with particular emphasis on small-and large-break loss-of-coolant-accidents (SB/LB-LOCAs) and main steam-line breaks. As a first step of this work, the best estimate system code TRACE has been used to develop a model of Olkiluoto-3. In order to test the nodalization, a scaling calculation from the rig of safety assessment (ROSA) test facility has been performed.
Inspections of existing nuclear power plants have pointed out the possibility that small break loss-of-coolant accidents (SBLOCAs) could be initiated by a small break located in the upper head of the reactor pressure vessel. Such type of breaks has been the subject of investigation in some of the tests carried out in the framework of the OECD/NEA ROSA test program for safety research and safety assessment of light water reactors. The ROSA/LSTF test facility simulates a Westinghouse design PWR with a four-loop configuration and 3423 MWth. The present paper is focused on Test 6-1 that simulated an RPV upper head SBLOCA with a break size equivalent to 1.9% cold leg break. The experiment assumes a total failure of the high pressure injection system and a loss of off-site power concurrent with the scram. The main purpose of the present study is the assessment of the capabilities of the best estimate system code, TRACE, to reproduce and understand the physical phenomena involved in this type of SBLOCA scenarios.
After the Specialist Meeting on Boron Dilution Reactivity Transients held in State College (USA) in 1995 and organized by the OECD in mutual effort with the NEA and the CSNI, one of the main concerns of the international thermal-hydraulic community was to analyse the capabilities of thermal-hydraulics codes to correctly simulate the phenomena involved in boron dilution transients. Small-break loss-of-coolant accidents (SB-LOCA) enable the formation of low-borated slugs in the loop seals. Low-borated water, if driven to the core, could cause a reactivity excursion. The work brought out in the present P.hd Thesis is mainly based on the participation in both the OECD/SETH and OECD/PKL projects that consisted among other issues on the experimental and simulation studies of SBLOCA with boron dilution transients in the PKL test facility. The participation of the Technical University of Catalonia (UPC) started with the development of a new nodalization with RELAP5 system code to simulate the PKL experiments. The nodalization proved its capability to simulate most of the involved phenomena. Findings and guidance about nodalization are pointed out. Transport of a low-borated slug through the primary system requires accuracy of the methods. Several studies have shown up high numerical diffusion introduced by the upwind difference schemes habitually used by system codes. The RELAP5 available models for the tracking of the boron concentration were studied and a new model was implemented to include physical diffusion in the boron mass conservation equation. The model was tested in all post-tests yielding satisfactory results. All lessons learned were afterwards applied for calculations on the Ascó NPP. A scaled calculation from one of the PKL tests was performed with satisfactory results that assured the capacity of the nodalization to reproduce the involved phenomena at least with the same accuracy as with the PKL nodalization. Thereafter, a base case and some sensitivity studies were carried out on the NPP nodalization.