Análisis de Estimación Óptima del Accidente de Pérdida de Refrigerante por Rotura Pequeña de 1”en Cabeza de Vasija en un Reactor tipo PWR

En el año 2002 durante una inspección se localizó una importante corrosión en la cabeza de la vasija de Davis Besse NPP. Si no se hubiera producido esa detección temprana, la corrosión hubiera provocado una pequeña rotura en la cabeza de la vasija. La OECD/NEA consideró la importancia de simular esta secuencia en la instalación experimental ROSA, la cual fue reproducida posteriormente por grupos de investigación internacionales con varios códigos de planta. En este caso el código utilizado para la simulación de las secuencias experimentales es TRACE. Los resultados de este test experimental fueron muy analizados internacionalmente por la gran influencia que dos factores tenía sobre el resultado: las acciones del operador relativas a la despresurización y la detección del descubrimiento del núcleo por los termopares que se encuentran a su salida. El comienzo del inicio de la despresurización del secundario estaba basado en la determinación del descubrimiento del núcleo por la lectura de los temopares de salida del núcleo. En el experimento se registró un retraso importante en la determinación de ese descubrimiento, comenzando la despresurización excesivamente tarde y haciendo necesaria la desactivación de los calentadores que simulan el núcleo del reactor para evitar su daño. Dada las condiciones excesivamente conservadoras del test experimentale, como el fallo de los dos trenes de inyección de alta presión durante todo el transitorio, en las aplicaciones de los experimentos con modelo de Almaraz NPP, se ha optado por reproducir dicho accidente con condiciones más realistas, verificando el impacto en los resultados de la disponibilidad de los trenes de inyección de alta presión o los tiempos de las acciones manuales del operador, como factores más limitantes y estableciendo el diámetro de rotura en 1”

Análisis de acciones del operador en un SBLOCA de 1”en el fondo de la Vasija en un Reactor tipo PWR

La simulación de accidentes de rotura pequeña en el fondo de la vasija se aparta del convencional análisis de LOCA de rama fría, el más limitante en los análisis deterministas La rotura de una de las penetraciones de instrumentación de la vasija ha sido desestimada históricamente en los análisis de licencia y en los Análisis Probabilistas de Seguridad y por ello, hay una falta evidente de literatura para dicho análisis. En el año 2003 durante una inspección, se detectó una considerable corrosión en el fondo de la vasija de South Texas Project Unit I NPP. La evolución en el tiempo de dicha corrosión habría derivado en una pequeña rotura en el fondo de la vasija si su detección no se hubiera producido a tiempo. La OECD/NEA consideró la importancia de simular dicha secuencia en la instalación experimental ROSA, la cual fue reproducida posteriormente por grupos de investigación internacionales con varios códigos de planta. En este caso el código utilizado para la simulación de las secuencias experimentales es TRACE. Tanto en el experimento como en la simulación se observaron las dificultades de reinundar la vasija al tener la rotura en el fondo de la misma, haciendo clave la gestión del accidente por parte del operador. Dadas las condiciones excesivamente conservadoras del test experimental, como el fallo de los dos trenes de inyección de alta presión durante todo el transitorio, en las aplicaciones de los experimentos con modelo de Almaraz NPP, se ha optado por reproducir dicho accidente con condiciones más realistas, verificando el impacto en los resultados de la disponibilidad de los trenes de inyección de alta presión o los tiempos de las acciones manuales del operador, como factores más limitantes y estableciendo el diámetro de rotura en 1”

Studies of a Self Cooled Lead Lithium blanket for HiPER Reactor

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates

A simplified Probabilistic Safety Assesment of a Steam-Methane Reforming Hydrogen Production Plant coupled to a High-Temperature Gas Cooled Nuclear Reactor

A Probabilistic Safety Assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a High-Temperature Gas Cooled Nuclear Reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute’s (JAERI) High Temperature Test Reactor (HTTR) prototype in Japan. This study has two major objectives: calculate the risk to onsite and offsite individuals, and calculate the frequency of different types of damage to the complex. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The initiating events presented here are methane pipe break, helium pipe break, and PPWC heat exchanger pipe break. Generic data was used for the fault tree analysis and the initiating event frequency. Saphire was used for the PSA analysis. The results show that the average frequency of an accident at this complex is 2.5E-06, which is divided into the various end states. The dominant sequences result in graphite oxidation which does not pose a health risk to the population. The dominant sequences that could affect the population are those that result in a methane explosion and occur 6.6E-8/year, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR. Sensitivity studies are being performed in order to determine where additional and improved data is needed.

Análisis Integrado de Seguridad de un accidente de SGTR en un reactor nuclear tipo PWR

El accidente de rotura de tubos de un generador de vapor (Steam Generator Tube Rupture, SGTR) en los reactores de agua a presión es uno de los transitorios más exigentes desde el punto de vista de operación. Los transitorios de SGTR son especiales, ya que podría dar lugar a emisiones radiológicas al exterior sin necesidad de daño en el núcleo previo o sin que falle la contención, ya que los SG pueden constituir una vía directa desde el reactor al medio ambiente en este transitorio. En los análisis de seguridad, el SGTR se analiza desde un punto determinista y probabilista, con distintos enfoques con respecto a las acciones del operador y las consecuencias analizadas. Cuando comenzaron los Análisis Deterministas de Seguridad (DSA), la forma de analizar el SGTR fue sin dar crédito a la acción del operador durante los primeros 30 min del transitorio, lo que suponía que el grupo de operación era capaz de detener la fuga por el tubo roto dentro de ese tiempo. Sin embargo, los diferentes casos reales de accidentes de SGTR sucedidos en los EE.UU. y alrededor del mundo demostraron que los operadores pueden emplear más de 30 minutos para detener la fuga en la vida real. Algunas metodologías fueron desarrolladas en los EEUU y en Europa para abordar esa cuestión. En el Análisis Probabilista de Seguridad (PSA), las acciones del operador se tienen en cuenta para diseñar los cabeceros en el árbol de sucesos. Los tiempos disponibles se utilizan para establecer los criterios de éxito para dichos cabeceros. Sin embargo, en una secuencia dinámica como el SGTR, las acciones de un operador son muy dependientes del tiempo disponible por las acciones humanas anteriores. Además, algunas de las secuencias de SGTR puede conducir a la liberación de actividad radiológica al exterior sin daño previo en el núcleo y que no se tienen en cuenta en el APS, ya que desde el punto de vista de la integridad de núcleo son de éxito. Para ello, para analizar todos estos factores, la forma adecuada de analizar este tipo de secuencias pueden ser a través de una metodología que contemple Árboles de Sucesos Dinámicos (Dynamic Event Trees, DET). En esta Tesis Doctoral se compara el impacto en la evolución temporal y la dosis al exterior de la hipótesis más relevantes encontradas en los Análisis Deterministas a nivel mundial. La comparación se realiza con un modelo PWR Westinghouse de tres lazos (CN Almaraz) con el código termohidráulico TRACE, con hipótesis de estimación óptima, pero con hipótesis deterministas como criterio de fallo único o pérdida de energía eléctrica exterior. Las dosis al exterior se calculan con RADTRAD, ya que es uno de los códigos utilizados normalmente para los cálculos de dosis del SGTR. El comportamiento del reactor y las dosis al exterior son muy diversas, según las diferentes hipótesis en cada metodología. Por otra parte, los resultados están bastante lejos de los límites de regulación, pese a los conservadurismos introducidos. En el siguiente paso de la Tesis Doctoral, se ha realizado un análisis de seguridad integrado del SGTR según la metodología ISA, desarrollada por el Consejo de Seguridad Nuclear español (CSN). Para ello, se ha realizado un análisis termo-hidráulico con un modelo de PWR Westinghouse de 3 lazos con el código MAAP. La metodología ISA permite la obtención del árbol de eventos dinámico del SGTR, teniendo en cuenta las incertidumbres en los tiempos de actuación del operador. Las simulaciones se realizaron con SCAIS (sistema de simulación de códigos para la evaluación de la seguridad integrada), que incluye un acoplamiento dinámico con MAAP. Las dosis al exterior se calcularon también con RADTRAD. En los resultados, se han tenido en cuenta, por primera vez en la literatura, las consecuencias de las secuencias en términos no sólo de daños en el núcleo sino de dosis al exterior. Esta tesis doctoral demuestra la necesidad de analizar todas las consecuencias que contribuyen al riesgo en un accidente como el SGTR. Para ello se ha hecho uso de una metodología integrada como ISA-CSN. Con este enfoque, la visión del DSA del SGTR (consecuencias radiológicas) se une con la visión del PSA del SGTR (consecuencias de daño al núcleo) para evaluar el riesgo total del accidente. Abstract Steam Generator Tube Rupture accidents in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are special transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path to the environment. The SGTR is analyzed from a Deterministic and Probabilistic point of view in the Safety Analysis, although the assumptions of the different approaches regarding the operator actions are quite different. In the beginning of Deterministic Safety Analysis, the way of analyzing the SGTR was not crediting the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that operators can took more than 30 min to stop the leakage in actual sequences. Some methodologies were raised in the USA and in Europe to cover that issue. In the Probabilistic Safety Analysis, the operator actions are taken into account to set the headers in the event tree. The available times are used to establish the success criteria for the headers. However, in such a dynamic sequence as SGTR, the operator actions are very dependent on the time available left by the other human actions. Moreover, some of the SGTR sequences can lead to offsite doses without previous core damage and they are not taken into account in PSA as from the point of view of core integrity are successful. Therefore, to analyze all this factors, the appropriate way of analyzing that kind of sequences could be through a Dynamic Event Tree methodology. This Thesis compares the impact on transient evolution and the offsite dose of the most relevant hypothesis of the different SGTR analysis included in the Deterministic Safety Analysis. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP), with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The offsite doses are calculated with RADTRAD code, as it is one of the codes normally used for SGTR offsite dose calculations. The behaviour of the reactor and the offsite doses are quite diverse depending on the different assumptions made in each methodology. On the other hand, although the high conservatism, such as the single failure criteria, the results are quite far from the regulatory limits. In the next stage of the Thesis, the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermohydraulical analysis of a Westinghouse 3-loop PWR plant with the MAAP code. The ISA methodology allows obtaining the SGTR Dynamic Event Tree taking into account the uncertainties on the operator actuation times. Simulations are performed with SCAIS (Simulation Code system for Integrated Safety Assessment), which includes a dynamic coupling with MAAP thermal hydraulic code. The offsite doses are calculated also with RADTRAD. The results shows the consequences of the sequences in terms not only of core damage but of offsite doses. This Thesis shows the need of analyzing all the consequences in an accident such as SGTR. For that, an it has been used an integral methodology like ISA-CSN. With this approach, the DSA vision of the SGTR (radiological consequences) is joined with the PSA vision of the SGTR (core damage consequences) to measure the total risk of the accident.

Self Cooled Lead Lithium blanket and reactor for HiPER

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates

Neutronics and activation of the preliminary reacion chamber of HiPER reactor based in a SCLL blanket

Neutronics and activation of the preliminary reacion chamber of HiPER reactor based in a SCLL blanket

Plasma facing materials for fusion reactor applications

The lack of plasma facing materials (PFM) able to withstand the severe magnetiicffusiion radiation conditions expected in fusion reactors is the actual bottle In both fusions approaches energy is released in the form of kinetic energy of neck for fusion to becomes a reality.

On track for solar grade silicon through a siemens process-type laboratory reactor: operating conditions and energy savings

Polysilicon cost impacts significantly on the photovoltaics (PV) cost and on the energy payback time. Nowadays, the besetting production process is the so called Siemens process, polysilicon deposition by chemical vapor deposition (CVD) from Trichlorosilane. Polysilicon purification level for PV is to a certain extent less demanding that for microelectronics. At the Instituto de Energía Solar (IES) research on this subject is performed through a Siemens process-type laboratory reactor. Through the laboratory CVD prototype at the IES laboratories, valuable information about the phenomena involved in the polysilicon deposition process and the operating conditions is obtained. Polysilicon deposition by CVD is a complex process due to the big number of parameters involved. A study on the influence of temperature and inlet gas mixture composition on the polysilicon deposition growth rate, based on experimental experience, is shown. Moreover, CVD process accounts for the largest contribution to the energy consumption of the polysilicon production. In addition, radiation phenomenon is the major responsible for low energetic efficiency of the whole process. This work presents a model of radiation heat loss, and the theoretical calculations are confirmed experimentally through a prototype reactor at our disposal, yielding a valuable know-how for energy consumption reduction at industrial Siemens reactors.

A comparative study of Monte Carlo-coupled depletion codes applied to a Sodium Fast Reactor design loaded with minor actinides

inor actinides (MAs) transmutation is a main design objective of advanced nuclear systems such as generation IV Sodium Fast Reactors (SFRs). In advanced fuel cycles, MA contents in final high level waste packages are main contributors to short term heat production as well as to long-term radiotoxicity. Therefore, MA transmutation would have an impact on repository designs and would reduce the environment burden of nuclear energy. In order to predict such consequences Monte Carlo (MC) transport codes are used in reactor design tasks and they are important complements and references for routinely used deterministic computational tools. In this paper two promising Monte Carlo transport-coupled depletion codes, EVOLCODE and SERPENT, are used to examine the impact of MA burning strategies in a SFR core, 3600 MWth. The core concept proposal for MA loading in two configurations is the result of an optimization effort upon a preliminary reference design to reduce the reactivity insertion as a consequence of sodium voiding, one of the main concerns of this technology. The objective of this paper is double. Firstly, efficiencies of the two core configurations for MA transmutation are addressed and evaluated in terms of actinides mass changes and reactivity coefficients. Results are compared with those without MA loading. Secondly, a comparison of the two codes is provided. The discrepancies in the results are quantified and discussed.

The european project NURISP for nuclear reactor simulation

The NURISP project aims at developing the European NURESIM reference simulation platform [1] for nuclear reactor. A first version of NURESIM was delivered in 2008. 22 organizations from 14 European countries contribute to the further development of this platform. NURISP also includes a User’s Group (UG) whose members are not NURISP partners and come from the industrial nuclear sector or European and non-European R&D labs. Users can benefit from the use of the NURESIM platform, methods, results and modules and they provide concrete input and feedback on the use of these elements.

Design and analysis of helium Brayton power cycles for HiPER reactor

Helium Brayton cycles have been studied as power cycles for both fission and fusion reactors obtaining high thermal efficiency. This paper studies several technological schemes of helium Brayton cycles applied for the HiPER reactor proposal. Since HiPER integrates technologies available at short term, its working conditions results in a very low maximum temperature of the energy sources, something that limits the thermal performance of the cycle. The aim of this work is to analyze the potential of the helium Brayton cycles as power cycles for HiPER. Several helium Brayton cycle configurations have been investigated with the purpose of raising the cycle thermal efficiency under the working conditions of HiPER. The effects of inter-cooling and reheating have specifically been studied. Sensitivity analyses of the key cycle parameters and component performances on the maximum thermal efficiency have also been carried out. The addition of several inter-cooling stages in a helium Brayton cycle has allowed obtaining a maximum thermal efficiency of over 36%, and the inclusion of a reheating process may also yield an added increase of nearly 1 percentage point to reach 37%. These results confirm that helium Brayton cycles are to be considered among the power cycle candidates for HiPER.

Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK-CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.

Methodologies for an improved prediction of the isotopic content in high burnup samples. Application to Vandellos-II reactor core

Fuel cycles are designed with the aim of obtaining the highest amount of energy possible. Since higher burnup values are reached, it is necessary to improve our disposal designs, traditionally based on the conservative assumption that they contain fresh fuel. The criticality calculations involved must consider burnup by making the most of the experimental and computational capabilities developed, respectively, to measure and predict the isotopic content of the spent nuclear fuel. These high burnup scenarios encourage a review of the computational tools to find out possible weaknesses in the nuclear data libraries, in the methodologies applied and their applicability range. Experimental measurements of the spent nuclear fuel provide the perfect framework to benchmark the most well-known and established codes, both in the industry and academic research activity. For the present paper, SCALE 6.0/TRITON and MONTEBURNS 2.0 have been chosen to follow the isotopic content of four samples irradiated in the Spanish Vandellós-II pressurized water reactor up to burnup values ranging from 40 GWd/MTU to 75 GWd/MTU. By comparison with the experimental data reported for these samples, we can probe the applicability of these codes to deal with high burnup problems. We have developed new computational tools within MONTENBURNS 2.0. They make possible to handle an irradiation history that includes geometrical and positional changes of the samples within the reactor core. This paper describes the irradiation scenario against which the mentioned codes and our capabilities are to be benchmarked.

Neutronics and activation of the preliminary reaction chamber of HiPER reactor based in a SCLL blanket

The HiPER reactor design is exploring different reaction chambers. In this study, we tackle the neutronicsand activation studies of a preliminary reaction chamber based in the following technologies: unpro-tected dry wall for the First Wall, self-cooled lead lithium blanket, and independent low activation steelVacuum Vessel. The most critical free parameter in this stage is the blanket thickness, as a function ofthe6Li enrichment. After a parametric study, we select for study both a ?thin? and ?thick? blanket, with?high? and ?low?6Li enrichment respectively, to reach a TBR = 1.1. To help to make a choice, we com-pute, for both blanket options, in addition to the TBR, the energy amplification factor, the tritium partialpressure, the203Hg and210Po total activity in the LiPb loop, and the Vacuum Vessel thickness requiredto guarantee the reweldability during its lifetime. The thin blanket shows a superior performance in thesafety related issues and structural viability, but it operates at higher6Li enrichment. It is selected forfurther improvements. The Vacuum Vessel shows to be unviable in both cases, with the thickness varyingbetween 39 and 52 cm. Further chamber modifications, such as the introduction of a neutron reflector,are required to exploit the benefits of the thin blanket with a reasonable Vacuum Vessel.