Application of a dynamic event tree methodolgy to steam generator tube rupture sequences

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermo-hydraulical analysis of a Westinghouse 3-loop PWR plant by means of the dynamic event trees (DET) for Steam Generator Tube Rupture (SGTR) sequences. The ISA methodology allows obtaining the SGTR Dynamic Event Tree taking into account 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 results show the capability of the ISA methodology and SCAIS platform to obtain the DET of complex sequences.

Behaviour of dolomite, olivine and alumina as primary catalysts in air-steam gasification of sewage sludge

Sewage sludge gasification assays were performed in an atmospheric fluidised bed reactor using air and air–steam mixtures as the gasifying agents. Dolomite, olivine and alumina are three well known tar removal catalysts used in biomass gasification processing. However, little information is available regarding their performance in sewage sludge gasification. The aim of the current study was to learn about the influence of these three catalysts in the product distribution and tar production during sewage sludge gasification. To this end, a set of assays was performed in which the temperature (750–850 °C), the in-bed catalyst content (0, 10 and 15 wt.%) and the steam–biomass ratio (SB) in the range of 0–1 were varied with a constant equivalence ratio (ER) of 0.3. The results were compared to the results from gasification without a catalyst. We show that dolomite has the highest activity in tar elimination, followed by alumina and olivine. In addition to improving tar removal, the presence of water vapour and the catalysts increased the content of H2 in the gases by nearly 60%.

Full scale test for explosion water barriers in small cross-section galleries

Underground coal mines explosions generally arise from the inflammation of a methane/air mixture. This explosion can also generate a subsequent coal dust explosion. Traditionally such explosions have being fought eliminating one or several of the factors needed by the explosion to take place. Although several preventive measures are taken to prevent explosions, other measures should be considered to reduce the effects or even to extinguish the flame front. Unlike other protection methods that remove one or two of the explosion triangle elements, namely; the ignition source, the oxidizing agent and the fuel, explosion barriers removes all of them: reduces the quantity of coal in suspension, cools the flame front and the steam generated by vaporization removes the oxygen present in the flame. Passive water barriers are autonomous protection systems against explosions that reduce to a satisfactory safety level the effects of methane and/or flammable dust explosions. The barriers are activated by the pressure wave provoked in the explosion destroying the barrier troughs and producing a uniform dispersion of the extinguishing agent throughout the gallery section in quantity enough to extinguish the explosion flame. Full scale tests have been carried out in Polish Barbara experimental mine at GIG Central Mining Institute in order to determine the requirements and the optimal installation conditions of these devices for small sections galleries which are very frequent in the Spanish coal mines. Full scale tests results have been analyzed to understand the explosion timing and development, in order to assess on the use of water barriers in the typical small crosssection Spanish galleries. Several arrangements of water barriers have been designed and tested to verify the effectiveness of the explosion suppression in each case. The results obtained demonstrate the efficiency of the water barriers in stopping the flame front even with smaller amounts of water than those established by the European standard. According to the tests realized, water barriers activation times are between 0.52 s and 0.78 s and the flame propagation speed are between 75 m/s and 80 m/s. The maximum pressures (Pmax) obtained in the full scale tests have varied between 0.2 bar and 1.8 bar. Passive barriers protect effectively against the spread of the flame but cannot be used as a safeguard of the gallery between the ignition source and the first row of water troughs or bags, or even after them, as the pressure could remain high after them even if the flame front has been extinguished.

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.

Simulación del accidente de SGTR en un PWR-W con TRACE para distintas metodologías de análisis determinista de seguridad

A Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) can lead to an atmospheric release bypassing the containment via the secondary system and exiting though the Pressurized Operating Relief Valves of the affected Steam Generator. That is why SGTR historically have been treated in a special way in the different Deterministic Safety Analysis (DSA), focusing on the radioactive release more than the possibility of core damage, as it is done in the other Loss of Coolant Accidents(LOCAs).

SolarPaces 2013: Innovations on direct steam generation in linear fresnel collectors

Direct Steam Generation (DSG) in Linear Fresnel (LF) solar collectors is being consolidated as a feasible technology for Concentrating Solar Power (CSP) plants. The competitiveness of this technology relies on the following main features: water as heat transfer fluid (HTF) in Solar Field (SF), obtaining high superheated steam temperatures and pressures at turbine inlet (500ºC and 90 bar), no heat tracing required to avoid HTF freezing, no HTF degradation, no environmental impacts, any heat exchanger between SF and Balance Of Plant (BOP), and low cost installation and maintenance. Regarding to LF solar collectors, were recently developed as an alternative to Parabolic Trough Collector (PTC) technology. The main advantages of LF are: the reduced collector manufacturing cost and maintenance, linear mirrors shapes versus parabolic mirror, fixed receiver pipes (no ball joints reducing leaking for high pressures), lower susceptibility to wind damages, and light supporting structures allowing reduced driving devices. Companies as Novatec, Areva, Solar Euromed, etc., are investing in LF DSG technology and constructing different pilot plants to demonstrate the benefits and feasibility of this solution for defined locations and conditions (Puerto Errado 1 and 2 in Murcia Spain, Lidellin Newcastle Australia, Kogran Creek in South West Queensland Australia, Kimberlina in Bakersfield California USA, Llo Solar in Pyrénées France,Dhursar in India,etc). There are several critical decisions that must be taken in order to obtain a compromise and optimization between plant performance, cost, and durability. Some of these decisions go through the SF design: proper thermodynamic operational parameters, receiver material selection for high pressures, phase separators and recirculation pumps number and location, pipes distribution to reduce the amount of tubes (reducing possible leaks points and transient time, etc.), etc. Attending to these aspects, the correct design parameters selection and its correct assessment are the main target for designing DSG LF power plants. For this purpose in the recent few years some commercial software tools were developed to simulatesolar thermal power plants, the most focused on LF DSG design are Thermoflex and System Advisor Model (SAM). Once the simulation tool is selected,it is made the study of the proposed SFconfiguration that constitutes the main innovation of this work, and also a comparison with one of the most typical state-of-the-art configuration. The transient analysis must be simulated with high detail level, mainly in the BOP during start up, shut down, stand by, and partial loads are crucial, to obtain the annual plant performance. An innovative SF configurationwas proposed and analyzed to improve plant performance. Finally it was demonstrated thermal inertia and BOP regulation mode are critical points in low sun irradiation day plant behavior, impacting in annual performance depending on power plant location.

Analysis of the operator action and the single failure criteria in a SGTR sequence using best estimate assumptions with TRACE 5.0

Steam Generator Tube Rupture (SGTR) sequences in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are a special kind of transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path from the reactor coolant system to the environment. The first methodology used to perform the Deterministic Safety Analysis (DSA) of a SGTR did not credit 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 period of time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that the operators usually take more than 30 min to stop the leakage in actual sequences. Some methodologies were raised to overcome that fact, considering operator actions from the beginning of the transient, as it is done in Probabilistic Safety Analysis. This paper presents the results of comparing different assumptions regarding the single failure criteria and the operator action taken from the most common methodologies included in the different Deterministic Safety Analysis. One single failure criteria that has not been analysed previously in the literature is proposed and analysed in this paper too. 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 behaviour of the reactor is quite diverse depending on the different assumptions made regarding the operator actions. On the other hand, although there are high conservatisms included in the hypothesis, as the single failure criteria, all the results are quite far from the regulatory limits. In addition, some improvements to the Emergency Operating Procedures to minimize the offsite release from the damaged SG in case of a SGTR are outlined taking into account the offsite dose sensitivity results.

Supercritical Steam power cycle for Line-Focus Solar Power Plants

The supercritical Rankine power cycle offers a net improvement in plant efficiency compared with a subcritical Rankine cycle. For fossil power plants the minimum supercritical steam turbine size is about 450MW. A recent study between Sandia National Laboratories and Siemens Energy, Inc., published on March 2013, confirmed the feasibility of adapting the Siemens turbine SST-900 for supercritical steam in concentrated solar power plants, with a live steam conditions 230-260 bar and output range between 140-200 MWe. In this context, this analysis is focused on integrating a line-focus solar field with a supercritical Rankine power cycle. For this purpose two heat transfer fluids were assessed: direct steam generation and molten salt Hitec XL. To isolate solar field from high pressure supercritical water power cycle, an intermediate heat exchanger was installed between linear solar collectors and balance of plant. Due to receiver selective coating temperature limitations, turbine inlet temperature was fixed 550ºC. The design-point conditions were 550ºC and 260 bar at turbine inlet, and 165 MWe Gross power output. Plant performance was assessed at design-point in the supercritical power plant (between 43-45% net plant efficiency depending on balance of plantconfiguration), and in the subcritical plant configuration (~40% net plant efficiency). Regarding the balance of plant configuration, direct reheating was adopted as the optimum solution to avoid any intermediate heat exchanger. One direct reheating stage between high pressure turbine and intermediate pressure turbine is the common practice; however, General Electric ultrasupercritical(350 bar) fossil power plants also considered doubled-reheat applications. In this study were analyzed heat balances with single-reheat, double-reheat and even three reheating stages. In all cases were adopted the proper reheating solar field configurations to limit solar collectors pressure drops. As main conclusion, it was confirmed net plant efficiency improvements in supercritical Rankine line-focus (parabolic or linear Fresnel) solar plant configurations are mainly due to the following two reasons: higher number of feed-water preheaters (up to seven)delivering hotter water at solar field inlet, and two or even three direct reheating stages (550ºC reheating temperature) in high or intermediate pressure turbines. However, the turbine manufacturer should confirm the equipment constrains regarding reheating stages and number of steam extractions to feed-water heaters.