[en] The pilot plant of fuel production of the National Institute of Nuclear Research (ININ) provided to the Laguna Verde Nuclear Power Plant (CNLV) four fuel assemblies type GE9B. The fuel irradiation was carried out in the unit 1 of the CNLV during four operation cycles, highlighting the fact that in their third cycle the four assemblies were placed in the center of the reactor core. In the Nuclear Systems Department (DSN) of the ININ it has been carried out studies to evaluate their neutron performance and to be able to determine the exposure levels of this fuels. Its also outlines the necessity to carry out a study of the thermomechanical behavior of the fuel rods that compose the assemblies, through computational codes that simulate their performance so much thermal as mechanical. For such purpose has been developing in the DSN the FETMA code, together with the codes that compose the system Fuel Management System (FMS), which evaluates the thermomechanical performance of fuel elements. In this work were used the FETMA and FEMAXI codes (developed by JAERI) to study the thermomechanical performance of the fuel elements manufactured in the ININ. (Author)

[en] The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

[en] In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

[en] The method of the response to the impulse of an autoregressive model for stability analysis of the nuclear boiling water reactors had one of the best behaviors in a range of stable operation conditions to quasi stables during the benchmark of stability of the Forsmark reactors. The method was developed in Mat lab and it uses the Gauss-Newton optimization method for to carry out the adjustment from the response to the impulse. In this work a program in Fortran of the response method to the impulse of an autoregressive model it was developed, which uses an adaptive optimization algorithm called NL2SOL, instead of the original method. This change is due that Gauss-Newton method doesn't converge in some cases to the best adjustment parameters for what the method has been substituted in the more recent Mat lab versions. Among the main obtained results it has that the programmed autoregressive model converges to a smaller order that the original method and while less stable is the reactor it is more big the difference in the order. Also was found an important difference in the first adjustment parameter being caused by the response magnitude to the impulse. As to for the decay ratio and oscillation frequency both programs presented acceptable results. (Author)

[en] The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

[en] The limitations in the burnt of the nuclear fuel usually are fixed by the one limit in the efforts to that undergo them the components of a nuclear fuel assembly. The limits defined its provide the direction to the fuel designer to reduce to the minimum the fuel failure during the operation, and they also prevent against some thermomechanical phenomena that could happen during the evolution of transitory events. Particularly, a limit value of LHGR is fixed to consider those physical phenomena that could lead to the interaction of the pellet-shirt (Pellet Cladding Interaction, PCI). This limit value it is related directly with an PCI limit that can be fixed based on experimental tests of power ramps. This way, to avoid to violate the PCI limit, the conditioning procedures of the fuel are still required for fuel elements with and without barrier. Those simulation procedures of the power ramp are carried out for the reactor operator during the starting maneuvers or of power increase like preventive measure of possible consequences in the thermomechanical behavior of the fuel. In this work, the thermomechanical behavior of two different types of fuel rods of the boiling water reactor is analyzed during the pursuit of the procedures of fuel preconditioning. Five diverse preconditioning calculations were carried out, each one with three diverse linear ramps of power increments. The starting point of the ramps was taken of the data of the cycle 8 of the unit 1 of the Laguna Verde Nucleo electric Central. The superior limit superior of the ramps it was the threshold of the lineal power in which a fuel failure could be presented by PCI, in function of the fuel burnt. The analysis was carried out with the FEMAXI-V code. (Author)

[en] The filtered venting of the containment has been adopted in European countries to mitigate the consequences of excess pressure containment during a severe accident. When venting has taken place, the fission products are released directly into the environment, unless a filter on the same path is placed, so that various types of filters are used to trap the fission products. The venting filters of the containment currently installed use different filtration technologies that involve more than one medium. Those using water as the first stage of filtration are called wet systems and are equipped with additional steps to remove water drops and fine aerosols emissions. And even they may also be equipped with an element containing certain absorption means for the filtration of gaseous iodine species. Other designs based on filtration of deep bed as the primary retention step; called dry filters, use filtration media of metal fiber, ceramic or sand to trap aerosols. This paper evaluates the hydraulic characteristics of the filter sand bed type designed by EDF as a candidate to be installed in the containment of BWR Mark II (type of primary containment of the nuclear power plant of Laguna Verde). The evaluation of filter sand bed type was performed using the software package of open source OpenFOAM. Models of each zone of the filtered device were generated and through a series of parametric calculations of computational fluid mechanics relevant hydrodynamic characteristics of the device were obtained, such as pressure drops against mass flow rate and pressure fields and speed at different operating conditions. On the other hand, the model validation of the sand bed filter when comparing the results of experimental tests on a sand column of PITEAS program (1985-1986) against OpenFOAM simulation was realized. The results are very close to those obtained experimentally. (Author)

[en] In this work was realized the simulation of the discharge transitory of both recirculation pumps of a BWR with the Simulate-3K code. This type of transitory is used in the stability analyses for the licensing of the fuel reload. An analysis of the precision of the simulation is also presented, using the FFTBM method jointly with the power relative contribution. This way, instead of determining the total precision of the calculation, a weighed precision is obtained by the contribution of each relevant parameter of the transitory. The results show that the precision of the simulation is acceptable due to the small magnitude of the merit figure of the width total average. The error in the merit figure comes mainly from the parameters total flow in the core and temperature of the fuel in the core. (Author)

[en] The filtered venting of the containment has been adopted in European countries to mitigate the consequences derived from the excess pressure of the containment during a severe accident. When venting has taken place, the fission products are released directly into the environment, unless a filter is placed in the path of the same, so various types of filters are used to trap the fission products. The containment venting filters currently installed use different filtering technologies that involve more than one medium. Those who use water as the first stage of filtration are called wet systems, are equipped with additional stages to eliminate water drops and emissions of fine aerosols, and may even be equipped with an element that contains certain means of absorption for the gaseous iodine species filtration. Other designs, based on deep bed filtration as the main retention stage, called dry filters; use metal fiber, ceramic or sand filtration media to trap aerosols. The present work evaluates the hydraulic characteristics of the sand bed type filter designed by EDF as a candidate to be installed in the containment of the BWR Mark II (primary containment type of the Laguna Verde nuclear power plant). The evaluation of the sand bed filter was performed using the OpenFOAM open source software package. Models of each zone of the filtering device were generated and by means of a series of parametric calculations of computational fluid mechanics, the relevant hydrodynamic characteristics of the device were obtained, such as pressure drops against mass flow and pressure fields and velocity under different operating conditions. On the other hand, the validation of the sand bed filter model was made when comparing the results of experimental tests carried out in a sand column of the PITEAS program (1985-1986) against the simulation in OpenFOAM. The results obtained are very close to those obtained experimentally. (Author)

[en] In this paper, several aspects are discussed about the implementation of an alternative source term for the analysis of the radiological consequences of design basis accidents in nuclear power plants. First, the rationale for implementation of an alternative source term is discussed. Then, the topics studied start by considering the current methodology and regulation applied to determine the original source term. Next, to determine a different source term, the basis of a new methodology is discussed, as, for example the elimination of excessive conservative assumptions. As a consequence of the adoption of an alternative source term, operational benefits are expected from relaxation of regulatory requirements established in the plant technical specifications. Other key issues considered in this work are the use of engineered safety features to minimize the iodine release during an accident, and technical requirements regarding the safe operation of the emergency filtering system for the main control room, in order to protect the reactor operation personnel. Finally, a discussion is presented about the impact on risk assessment, when using an alternative source term, and remarking that the adoption of a new source term by itself do not have and impact on plant risk, but it does have an effect on radiological consequences. Nevertheless, a detailed review of technical specification changes that could induce some risk should be considered. As conclusions of this work, recommendations are presented for the licensing process of an alternative source term. (Author)