Dynamic response of hydro power plants for providing secondary regulation reserves considering elastic water column effects

In this paper, the dynamic response of a hydro power plant for providing secondary regulation reserve is studied in detail. Special emphasis is given to the elastic water column effects both in the penstock and the tailrace tunnel. For this purpose, a nonlinear model based on the analogy between mass and momentum conservation equations of a water conduit and those of wave propagation in transmission lines is used. The influence of the plant configuration and design parameters on the fulfilment of the Spanish Electrical System Operator requirements is analysed

Analysis of the Behaviour of Biofuel-Fired Gas Turbine Power Plants

The utilisation of biofuels in gas turbines is a promising alternative to fossil fuels for power generation. It would lead to significant reduction of CO2 emissions using an existing combustion technology, although significant changes seem to be needed and further technological development is necessary. The goal of this work is to perform energy and exergy analyses of the behaviour of gas turbines fired with biogas, ethanol and synthesis gas (bio-syngas), compared with natural gas. The global energy transformation process (i.e. from biomass to electricity) has also been studied. Furthermore, the potential reduction of CO2 emissions attained by the use of biofuels has been determined, considering the restrictions regarding biomass availability. Two different simulation tools have been used to accomplish the aims of this work. The results suggest a high interest and the technical viability of the use of Biomass Integrated Gasification Combined Cycle (BIGCC) systems for large scale power generation.

Dynamic response of hydro power plants to load variations for providing secondary regulation reserves considering elastic water column effects

In this paper, the dynamic response of a hydro power plant for providing secondary regulation reserve is studied in detail. S pecial emphasis is given to the elastic water column effects both in the penstock and the tailrace tunnel. For this purpose, a nonline ar model based on the analogy between mass and momentum conservation equations of a water conduit and those of wave propagation in transmission lines is used. The influence of the plant configuration and design parameters on the fulfilment of the Spanish Electrical System Operator requirem ents is analysed.

Testing a GIS-based methodology for optimal location of Pumped Storage power plants in Norway

Europe needs to restructure its energy system. The aim to decrease the reliance on fossil fuels to a higher dependence on renewable energy has now been imposed by The European Commission. In order to achieve this goal there is a great interest in Norway to become "The Green Battery of Europe". In the pursuit of this goal a GIS-tool was created to investigate the pump storage potential in Norway. The tool searches for possible connections between existing reservoirs and dams with the criteria selected by the user. The aim of this thesis was to test the tool and see if the results suggested were plausible, develop a cost calculation method for the PSH lines, and make suggestions for further development of the tool. During the process the tool presented many non-feasible pumped storage hydropower (PSH) connections. The area of Telemark was chosen for the more detailed study. The results were discussed and some improvements were suggested for further development of the tool. Also a sensitivity test was done to see which of the parameters set by the user are the most relevant for the PSH connection suggestion. From a range of the most promising PSH plants suggested by the tool, the one between Songavatn and Totak was chosen for a case study, where there already exists a power plant between both reservoirs. A new Pumped Storage Plant was designed with a power production of 1200 MW. There are still many topics open to discussion, such as how to deal with environmental restrictions, or how to deal with inflows and outflows of the reservoirs from the existing power plants. Consequently the GIS-tool can be a very useful tool to establish the best possible connections between existing reservoirs and dams, but it still needs a deep study and the creation of new parameters for the user.

Dry cooling with night cool storage to enhance solar power plants performance in extreme conditions areas

Solar thermal power plants are usually installed in locations with high yearly average solar radiation, often deserts. In such conditions, cooling water required for thermodynamic cycles is rarely available. Moreover, when solar radiation is high, ambient temperature is very high as well; this leads to excessive condensation temperature, especially when air-condensers are used, and decreases the plant efficiency. However, temperature variation in deserts is often very high, which drives to relatively low temperatures during the night. This fact can be exploited with the use of a closed cooling system, so that the coolant (water) is chilled during the night and store. Chilled water is then used during peak temperature hours to cool the condenser (dry cooling), thus enhancing power output and efficiency. The present work analyzes the performance improvement achieved by night thermal cool storage, compared to its equivalent air cooled power plant. Dry cooling is proved to be energy-effective for moderately high day–night temperature differences (20 °C), often found in desert locations. The storage volume requirement for different power plant efficiencies has also been studied, resulting on an asymptotic tendency.

Comparison between SAM and Thermoflow for Linear Fresnel Solar Power Plants design

System Advisor Model is a software tool develped by National Renewable Laboratory (NREL), Department Of Energy, USA to design Solar Power Plants.

Co2 capture in power plants-using the oxy-combustion principle

In the C02 capture from power generation, the energy penalties for the capture are one of the main challenges. Nowadays, the post-combustion methods have energy penalties 10wer than the oxy combustion and pre-combustion technologies. One of the main disadvantages of the post combustion method is the fact that the capture ofC02at atmospheric pressure requires quite big equipment for the high flow rates of flue gas, and the 10w partial pressure of the CO2generates an important 10ss of energy. The A1lam cyc1e presented for NETPOWER gives high efficiencies in the power production and 10w energy penalties. A simulation of this cyc1e is made together with a simulation of power plants with pre-combustion and post-combustion capture and without capture for natural gas and forcoa1. The simulations give 10wer efficiencies than the proposed for NETPOWER For natural gas the efficiency is 52% instead of the 59% presented, and 33% instead of51% in the case of using coal as fuel. Are brought to light problems in the CO2compressor due the high flow ofC02that is compressed unti1300 bar to be recyc1ed into the combustor.

Overview of the Spanish Fuel Cycle: Technical Tours Organized by Spanish young Generation in Nuclear.

Spanish Young Generation in Nuclear (Jóvenes Nucleares, JJNN) is a non-profrt organization that depends on the Spanish Nuclear Society (Sociedad Nuclear Española, SNE).Since one of rts main goals is to spread the knowledge about nuclear power,severa! technical tours to facilities wrth an importan!role in the nuclear fuel cycle have been organized for the purpose ofleaming about the different stages of the Spanish tuel cycle. Spanish Young Generation in Nuclear had the opportunity to visit ENUSA Fuel Assembly Factory in Juzbado (Salamanca, Spain), Where it could be understood the front-end cycle which involves the uranium supply and storage, design and manufacturing of fuel bundles for European nuclear power plants. Alterwards, due to the tour of Almaraz NPP (PWR) and Santa María de Garoña NPP (BWR), rt could be comprehended how to obtain energy from this fuel in two different types of reactors.Furthermore,in these two plants, the facilities related to the back-end cycle could be toured. lt was possible to watch the Spent FuelPools, where the fuel bundles are stored under water until their activity is reduced enough to transport them to an Individual Temporary Storage Facility orto the Centralized Temporary Storage. Finally, a technical tour to ENSA Heavy Components Factory (ENSA) was accomplished, Where it could be experienced at first hand how different Nuclear Steam Supply System (NSSS) components and other nuclear elements, such as racks or shipping and storage casks for spent nuclear fuel, are manulactured. All these perlonned technical tours were a complete success thanks to a generous care and know-how of the wor1

Characteristic study, its identification and self-tuned approach to control hydro-power plants

The water time constant and mechanical time constant greatly influences the power and speed oscillations of hydro-turbine-generator unit. This paper discusses the turbine power transients in response to different nature and changes in the gate position. The work presented here analyses the characteristics of hydraulic system with an emphasis on changes in the above time constants. The simulation study is based on mathematical first-, second-, third- and fourth-order transfer function models. The study is further extended to identify discrete time-domain models and their characteristic representation without noise and with noise content of 10 & 20 dB signal-to-noise ratio (SNR). The use of self-tuned control approach in minimising the speed deviation under plant parameter changes and disturbances is also discussed.

Development of a methodology for nuclear data uncertainty propagation on isotopic evolution calculations for advanced nuclear systems

Una apropiada evaluación de los márgenes de seguridad de una instalación nuclear, por ejemplo, una central nuclear, tiene en cuenta todas las incertidumbres que afectan a los cálculos de diseño, funcionanmiento y respuesta ante accidentes de dicha instalación. Una fuente de incertidumbre son los datos nucleares, que afectan a los cálculos neutrónicos, de quemado de combustible o activación de materiales. Estos cálculos permiten la evaluación de las funciones respuesta esenciales para el funcionamiento correcto durante operación, y también durante accidente. Ejemplos de esas respuestas son el factor de multiplicación neutrónica o el calor residual después del disparo del reactor. Por tanto, es necesario evaluar el impacto de dichas incertidumbres en estos cálculos. Para poder realizar los cálculos de propagación de incertidumbres, es necesario implementar metodologías que sean capaces de evaluar el impacto de las incertidumbres de estos datos nucleares. Pero también es necesario conocer los datos de incertidumbres disponibles para ser capaces de manejarlos. Actualmente, se están invirtiendo grandes esfuerzos en mejorar la capacidad de analizar, manejar y producir datos de incertidumbres, en especial para isótopos importantes en reactores avanzados. A su vez, nuevos programas/códigos están siendo desarrollados e implementados para poder usar dichos datos y analizar su impacto. Todos estos puntos son parte de los objetivos del proyecto europeo ANDES, el cual ha dado el marco de trabajo para el desarrollo de esta tesis doctoral. Por tanto, primero se ha llevado a cabo una revisión del estado del arte de los datos nucleares y sus incertidumbres, centrándose en los tres tipos de datos: de decaimiento, de rendimientos de fisión y de secciones eficaces. A su vez, se ha realizado una revisión del estado del arte de las metodologías para la propagación de incertidumbre de estos datos nucleares. Dentro del Departamento de Ingeniería Nuclear (DIN) se propuso una metodología para la propagación de incertidumbres en cálculos de evolución isotópica, el Método Híbrido. Esta metodología se ha tomado como punto de partida para esta tesis, implementando y desarrollando dicha metodología, así como extendiendo sus capacidades. Se han analizado sus ventajas, inconvenientes y limitaciones. El Método Híbrido se utiliza en conjunto con el código de evolución isotópica ACAB, y se basa en el muestreo por Monte Carlo de los datos nucleares con incertidumbre. En esta metodología, se presentan diferentes aproximaciones según la estructura de grupos de energía de las secciones eficaces: en un grupo, en un grupo con muestreo correlacionado y en multigrupos. Se han desarrollado diferentes secuencias para usar distintas librerías de datos nucleares almacenadas en diferentes formatos: ENDF-6 (para las librerías evaluadas), COVERX (para las librerías en multigrupos de SCALE) y EAF (para las librerías de activación). Gracias a la revisión del estado del arte de los datos nucleares de los rendimientos de fisión se ha identificado la falta de una información sobre sus incertidumbres, en concreto, de matrices de covarianza completas. Además, visto el renovado interés por parte de la comunidad internacional, a través del grupo de trabajo internacional de cooperación para evaluación de datos nucleares (WPEC) dedicado a la evaluación de las necesidades de mejora de datos nucleares mediante el subgrupo 37 (SG37), se ha llevado a cabo una revisión de las metodologías para generar datos de covarianza. Se ha seleccionando la actualización Bayesiana/GLS para su implementación, y de esta forma, dar una respuesta a dicha falta de matrices completas para rendimientos de fisión. Una vez que el Método Híbrido ha sido implementado, desarrollado y extendido, junto con la capacidad de generar matrices de covarianza completas para los rendimientos de fisión, se han estudiado diferentes aplicaciones nucleares. Primero, se estudia el calor residual tras un pulso de fisión, debido a su importancia para cualquier evento después de la parada/disparo del reactor. Además, se trata de un ejercicio claro para ver la importancia de las incertidumbres de datos de decaimiento y de rendimientos de fisión junto con las nuevas matrices completas de covarianza. Se han estudiado dos ciclos de combustible de reactores avanzados: el de la instalación europea para transmutación industrial (EFIT) y el del reactor rápido de sodio europeo (ESFR), en los cuales se han analizado el impacto de las incertidumbres de los datos nucleares en la composición isotópica, calor residual y radiotoxicidad. Se han utilizado diferentes librerías de datos nucleares en los estudios antreriores, comparando de esta forma el impacto de sus incertidumbres. A su vez, mediante dichos estudios, se han comparando las distintas aproximaciones del Método Híbrido y otras metodologías para la porpagación de incertidumbres de datos nucleares: Total Monte Carlo (TMC), desarrollada en NRG por A.J. Koning y D. Rochman, y NUDUNA, desarrollada en AREVA GmbH por O. Buss y A. Hoefer. Estas comparaciones demostrarán las ventajas del Método Híbrido, además de revelar sus limitaciones y su rango de aplicación. ABSTRACT For an adequate assessment of safety margins of nuclear facilities, e.g. nuclear power plants, it is necessary to consider all possible uncertainties that affect their design, performance and possible accidents. Nuclear data are a source of uncertainty that are involved in neutronics, fuel depletion and activation calculations. These calculations can predict critical response functions during operation and in the event of accident, such as decay heat and neutron multiplication factor. Thus, the impact of nuclear data uncertainties on these response functions needs to be addressed for a proper evaluation of the safety margins. Methodologies for performing uncertainty propagation calculations need to be implemented in order to analyse the impact of nuclear data uncertainties. Nevertheless, it is necessary to understand the current status of nuclear data and their uncertainties, in order to be able to handle this type of data. Great eórts are underway to enhance the European capability to analyse/process/produce covariance data, especially for isotopes which are of importance for advanced reactors. At the same time, new methodologies/codes are being developed and implemented for using and evaluating the impact of uncertainty data. These were the objectives of the European ANDES (Accurate Nuclear Data for nuclear Energy Sustainability) project, which provided a framework for the development of this PhD Thesis. Accordingly, first a review of the state-of-the-art of nuclear data and their uncertainties is conducted, focusing on the three kinds of data: decay, fission yields and cross sections. A review of the current methodologies for propagating nuclear data uncertainties is also performed. The Nuclear Engineering Department of UPM has proposed a methodology for propagating uncertainties in depletion calculations, the Hybrid Method, which has been taken as the starting point of this thesis. This methodology has been implemented, developed and extended, and its advantages, drawbacks and limitations have been analysed. It is used in conjunction with the ACAB depletion code, and is based on Monte Carlo sampling of variables with uncertainties. Different approaches are presented depending on cross section energy-structure: one-group, one-group with correlated sampling and multi-group. Differences and applicability criteria are presented. Sequences have been developed for using different nuclear data libraries in different storing-formats: ENDF-6 (for evaluated libraries) and COVERX (for multi-group libraries of SCALE), as well as EAF format (for activation libraries). A revision of the state-of-the-art of fission yield data shows inconsistencies in uncertainty data, specifically with regard to complete covariance matrices. Furthermore, the international community has expressed a renewed interest in the issue through the Working Party on International Nuclear Data Evaluation Co-operation (WPEC) with the Subgroup (SG37), which is dedicated to assessing the need to have complete nuclear data. This gives rise to this review of the state-of-the-art of methodologies for generating covariance data for fission yields. Bayesian/generalised least square (GLS) updating sequence has been selected and implemented to answer to this need. Once the Hybrid Method has been implemented, developed and extended, along with fission yield covariance generation capability, different applications are studied. The Fission Pulse Decay Heat problem is tackled first because of its importance during events after shutdown and because it is a clean exercise for showing the impact and importance of decay and fission yield data uncertainties in conjunction with the new covariance data. Two fuel cycles of advanced reactors are studied: the European Facility for Industrial Transmutation (EFIT) and the European Sodium Fast Reactor (ESFR), and response function uncertainties such as isotopic composition, decay heat and radiotoxicity are addressed. Different nuclear data libraries are used and compared. These applications serve as frameworks for comparing the different approaches of the Hybrid Method, and also for comparing with other methodologies: Total Monte Carlo (TMC), developed at NRG by A.J. Koning and D. Rochman, and NUDUNA, developed at AREVA GmbH by O. Buss and A. Hoefer. These comparisons reveal the advantages, limitations and the range of application of the Hybrid Method.

Sistema de simulación de escenarios de ciclos avanzados de combustible nuclear en transición = Simulation system for advanced fuel cycle transition scenarios

El estudio de los ciclos del combustible nuclear requieren de herramientas computacionales o "códigos" versátiles para dar respuestas al problema multicriterio de evaluar los actuales ciclos o las capacidades de las diferentes estrategias y escenarios con potencial de desarrollo en a nivel nacional, regional o mundial. Por otra parte, la introducción de nuevas tecnologías para reactores y procesos industriales hace que los códigos existentes requieran nuevas capacidades para evaluar la transición del estado actual del ciclo del combustible hacia otros más avanzados y sostenibles. Brevemente, esta tesis se centra en dar respuesta a las principales preguntas, en términos económicos y de recursos, al análisis de escenarios de ciclos de combustible, en particular, para el análisis de los diferentes escenarios del ciclo del combustible de relativa importancia para España y Europa. Para alcanzar este objetivo ha sido necesaria la actualización y el desarrollo de nuevas capacidades del código TR_EVOL (Transition Evolution code). Este trabajo ha sido desarrollado en el Programa de Innovación Nuclear del CIEMAT desde el año 2010. Esta tesis se divide en 6 capítulos. El primer capítulo ofrece una visión general del ciclo de combustible nuclear, sus principales etapas y los diferentes tipos utilizados en la actualidad o en desarrollo para el futuro. Además, se describen las fuentes de material nuclear que podrían ser utilizadas como combustible (uranio y otros). También se puntualizan brevemente una serie de herramientas desarrolladas para el estudio de estos ciclos de combustible nuclear. El capítulo 2 está dirigido a dar una idea básica acerca de los costes involucrados en la generación de electricidad mediante energía nuclear. Aquí se presentan una clasificación de estos costos y sus estimaciones, obtenidas en la bibliografía, y que han sido evaluadas y utilizadas en esta tesis. Se ha incluido también una breve descripción del principal indicador económico utilizado en esta tesis, el “coste nivelado de la electricidad”. El capítulo 3 se centra en la descripción del código de simulación desarrollado para el estudio del ciclo del combustible nuclear, TR_EVOL, que ha sido diseñado para evaluar diferentes opciones de ciclos de combustibles. En particular, pueden ser evaluados las diversos reactores con, posiblemente, diferentes tipos de combustibles y sus instalaciones del ciclo asociadas. El módulo de evaluaciones económica de TR_EVOL ofrece el coste nivelado de la electricidad haciendo uso de las cuatro fuentes principales de información económica y de la salida del balance de masas obtenido de la simulación del ciclo en TR_EVOL. Por otra parte, la estimación de las incertidumbres en los costes también puede ser efectuada por el código. Se ha efectuado un proceso de comprobación cruzada de las funcionalidades del código y se descrine en el Capítulo 4. El proceso se ha aplicado en cuatro etapas de acuerdo con las características más importantes de TR_EVOL, balance de masas, composición isotópica y análisis económico. Así, la primera etapa ha consistido en el balance masas del ciclo de combustible nuclear actual de España. La segunda etapa se ha centrado en la comprobación de la composición isotópica del flujo de masas mediante el la simulación del ciclo del combustible definido en el proyecto CP-ESFR UE. Las dos últimas etapas han tenido como objetivo validar el módulo económico. De este modo, en la tercera etapa han sido evaluados los tres principales costes (financieros, operación y mantenimiento y de combustible) y comparados con los obtenidos por el proyecto ARCAS, omitiendo los costes del fin del ciclo o Back-end, los que han sido evaluado solo en la cuarta etapa, haciendo uso de costes unitarios y parámetros obtenidos a partir de la bibliografía. En el capítulo 5 se analizan dos grupos de opciones del ciclo del combustible nuclear de relevante importancia, en términos económicos y de recursos, para España y Europa. Para el caso español, se han simulado dos grupos de escenarios del ciclo del combustible, incluyendo estrategias de reproceso y extensión de vida de los reactores. Este análisis se ha centrado en explorar las ventajas y desventajas de reprocesado de combustible irradiado en un país con una “relativa” pequeña cantidad de reactores nucleares. Para el grupo de Europa se han tratado cuatro escenarios, incluyendo opciones de transmutación. Los escenarios incluyen los reactores actuales utilizando la tecnología reactor de agua ligera y ciclo abierto, un reemplazo total de los reactores actuales con reactores rápidos que queman combustible U-Pu MOX y dos escenarios del ciclo del combustible con transmutación de actínidos minoritarios en una parte de los reactores rápidos o en sistemas impulsados por aceleradores dedicados a transmutación. Finalmente, el capítulo 6 da las principales conclusiones obtenidas de esta tesis y los trabajos futuros previstos en el campo del análisis de ciclos de combustible nuclear. ABSTRACT The study of the nuclear fuel cycle requires versatile computational tools or “codes” to provide answers to the multicriteria problem of assessing current nuclear fuel cycles or the capabilities of different strategies and scenarios with potential development in a country, region or at world level. Moreover, the introduction of new technologies for reactors and industrial processes makes the existing codes to require new capabilities to assess the transition from current status of the fuel cycle to the more advanced and sustainable ones. Briefly, this thesis is focused in providing answers to the main questions about resources and economics in fuel cycle scenario analyses, in particular for the analysis of different fuel cycle scenarios with relative importance for Spain and Europe. The upgrade and development of new capabilities of the TR_EVOL code (Transition Evolution code) has been necessary to achieve this goal. This work has been developed in the Nuclear Innovation Program at CIEMAT since year 2010. This thesis is divided in 6 chapters. The first one gives an overview of the nuclear fuel cycle, its main stages and types currently used or in development for the future. Besides the sources of nuclear material that could be used as fuel (uranium and others) are also viewed here. A number of tools developed for the study of these nuclear fuel cycles are also briefly described in this chapter. Chapter 2 is aimed to give a basic idea about the cost involved in the electricity generation by means of the nuclear energy. The main classification of these costs and their estimations given by bibliography, which have been evaluated in this thesis, are presented. A brief description of the Levelized Cost of Electricity, the principal economic indicator used in this thesis, has been also included. Chapter 3 is focused on the description of the simulation tool TR_EVOL developed for the study of the nuclear fuel cycle. TR_EVOL has been designed to evaluate different options for the fuel cycle scenario. In particular, diverse nuclear power plants, having possibly different types of fuels and the associated fuel cycle facilities can be assessed. The TR_EVOL module for economic assessments provides the Levelized Cost of Electricity making use of the TR_EVOL mass balance output and four main sources of economic information. Furthermore, uncertainties assessment can be also carried out by the code. A cross checking process of the performance of the code has been accomplished and it is shown in Chapter 4. The process has been applied in four stages according to the most important features of TR_EVOL. Thus, the first stage has involved the mass balance of the current Spanish nuclear fuel cycle. The second stage has been focused in the isotopic composition of the mass flow using the fuel cycle defined in the EU project CP-ESFR. The last two stages have been aimed to validate the economic module. In the third stage, the main three generation costs (financial cost, O&M and fuel cost) have been assessed and compared to those obtained by ARCAS project, omitting the back-end costs. This last cost has been evaluated alone in the fourth stage, making use of some unit cost and parameters obtained from the bibliography. In Chapter 5 two groups of nuclear fuel cycle options with relevant importance for Spain and Europe are analyzed in economic and resources terms. For the Spanish case, two groups of fuel cycle scenarios have been simulated including reprocessing strategies and life extension of the current reactor fleet. This analysis has been focused on exploring the advantages and disadvantages of spent fuel reprocessing in a country with relatively small amount of nuclear power plants. For the European group, four fuel cycle scenarios involving transmutation options have been addressed. Scenarios include the current fleet using Light Water Reactor technology and open fuel cycle, a full replacement of the initial fleet with Fast Reactors burning U-Pu MOX fuel and two fuel cycle scenarios with Minor Actinide transmutation in a fraction of the FR fleet or in dedicated Accelerator Driven Systems. Finally, Chapter 6 gives the main conclusions obtained from this thesis and the future work foreseen in the field of nuclear fuel cycle analysis.

Fresnel-based modular solar fields for performance/cost optimization in solar thermal power plants: A comparison with parabolic trough collectors.

Linear Fresnel collectors are identified as a technology that should play a main role in order to reduce cost of Concentrating Solar Power. An optical and thermal analysis of the different blocks of the solar power plant is carried out, where Fresnel arrays are compared with the most extended linear technology: parabolic trough collectors. It is demonstrated that the optical performance of Fresnel array is very close to that of PTC, with similar values of maximum flux intensities. In addition, if the heat carrier fluid flows in series by the tubes of the receiver, relatively high thermal efficiencies are achieved. Thus, an annual solar to electricity efficiency of 19% is expected, which is similar to the state of the art in PTCs; this is done with a reduction of costs, thanks to lighter structures, that drives to an estimation of LCOE of around 6.5 c€/kWh.

The use of geographic information systems for the optimal location of biomass power plants in the Madrid Community (Spain)

This article has been extracted from the results of a thesis entitled “Potential bioelectricity production of the Madrid Community Agricultural Regions based on rye and triticale biomass.” The aim was, first, to quantify the potential of rye (Secale Cereale L.) and triticale ( Triticosecale Aestivum L.) biomass in each of the Madrid Community agricultural regions, and second, to locate the most suitable areas for the installation of power plants using biomass. At least 17,339.9 t d.m. of rye and triticale would be required to satisfy the biomass needs of a 2.2 MW power plant, (considering an efficiency of 21.5%, 8,000 expected operating hours/year and a biomass LCP of 4,060 kcal/kg for both crops), and 2,577 ha would be used (which represent 2.79% of the Madrid Community fallow dry land surface). Biomass yields that could be achieved in Madrid Community using 50% of the fallow dry land surface (46,150 ha representing 5.75% of the Community area), based on rye and triticale crops, are estimated at 84,855, 74,906, 70,109, 50,791, 13,481, and 943 t annually for the Campiña, Vegas, Sur Occidental, Área Metropolitana, Lozoya-Somosierra, and Guadarrama regions. The latter represents a bioelectricity potential of 10.77, 9.5, 8.9, 6.44, 1.71, and 0.12 MW, respectively.