Racial/ethnic differences in factors influencing screening and treatment among prostate cancer patients

Background. Racial/ethnic differences have been found in various aspects of cancer care. But a limited number of studies have examined the racial/ethnic differences in predictors of prostate-specific antigen (PSA) screening in a group of prostate cancer patients and have attempted to identify the racial/ethnic differences in treatment discussions, treatment choice and treatment received for organ-confined localized prostate cancer (PCa) among three major racial/ethnic groups of the USA. This study was conducted to redress this lack of information. ^ Methods. This study was conducted on a group of 935 prostate cancer patients representing all three major race/ethnic groups (Whites, African Americans and Hispanics) who were treated at various medical institutes of the Texas Medical Center, Houston between 1996 and 2004 to identify the racial/ethnic differences in predictors of PSA screening. A subset of 640 patients who had organ-confined localized prostate cancer was selected to examine the racial/ethnic differences in treatment discussions, treatment choice and treatment received for their localized prostate cancer. They were interviewed by trained research interviewers of MD Anderson Cancer Center using a validated structured questionnaire. ^ Results. The results showed that African American (54.4%) and Hispanic patients (42.3%) were significantly less likely (p=0.004 and p<.001, respectively) than White patients (63.2%) to report having had PSA screening before their prostate-cancer diagnosis. Among Whites, only education and annual check-ups predicted the use of PSA screening, whereas in African Americans two more additional factors, marital status and bode-mass index (BMI), significantly predicted PSA screening. Among Hispanics, like two other groups, education and annual check-ups also appeared as a significant predictor of PSA screening. ^ Results from multivariable logistic regression showed that African American patients were 15% less likely (OR=0.85, 95% CI=0.61-1.17, p=0.32) and Hispanics patients were 40% less likely (OR=0.60, 95% CI=0.41-0.87, p=0.008) to undergo PSA screening than Whites after adjusting for education and age at diagnosis for African Americans, and for education, annual check-ups and age at diagnosis for Hispanics. ^ This study revealed that health professionals were less likely to discuss surgery (79.9% vs. 93.2%) and watchful waiting (27.9% vs. 43.9%) with Hispanics compared to Whites. African Americans were more likely to choose (35.1% vs. 27.7%) and receive radiation therapy (38.3% vs.31.4%) than Whites. A comparison of concordance between treatment choice and treatment received showed that the highest concordance was found for watchful waiting and radiation therapy among African Americans (100% and 85.9%, respectively) whereas the highest concordance (96.9%) was found for surgery among Hispanics. ^ Conclusions. In this multiethnic study, the rates of PSA screening and its potential predictors varied by racial/ethnic groups. Substantial racial/ethnic variations were also found in treatment discussion, but the differences were not evident for treatment choice and treatment received. Health-education programs and culturally appropriate educational outreach efforts, especially targeted for high-risk groups, are needed to reduce these disparities. In the current climate of uncertainty about the benefits of PSA screening, or the benefit of one treatment over others, men should have access to information and services regardless of race/ethnicity so that they can make informed decisions. Further in-depth studies are needed in other settings to confirm these findings with the goal of developing an intervention to address these concerns. ^

Affect and cognition measures in preference-based decisions: Validity testing of the Ottawa Decisional Conflict Scale and a decision-specific anxiety measure with men eligible for prostate cancer screening

Background. At present, prostate cancer screening (PCS) guidelines require a discussion of risks, benefits, alternatives, and personal values, making decision aids an important tool to help convey information and to help clarify values. Objective: The overall goal of this study is to provide evidence of the reliability and validity of a PCS anxiety measure and the Decisional Conflict Scale (DCS). Methods. Using data from a randomized, controlled PCS decision aid trial that measured PCS anxiety at baseline and DCS at baseline (T0) and at two-weeks (T2), four psychometric properties were assessed: (1) internal consistency reliability, indicated by factor analysis intraclass correlations and Cronbach's α; (2) construct validity, indicated by patterns of Pearson correlations among subscales; (3) discriminant validity, indicated by the measure's ability to discriminate between undecided men and those with a definite screening intention; and (4) factor validity and invariance using confirmatory factor analyses (CFA). Results. The PCS anxiety measure had adequate internal consistency reliability and good construct and discriminant validity. CFAs indicated that the 3-factor model did not have adequate fit. CFAs for a general PCS anxiety measure and a PSA anxiety measure indicated adequate fit. The general PCS anxiety measure was invariant across clinics. The DCS had adequate internal consistency reliability except for the support subscale and had adequate discriminate validity. Good construct validity was found at the private clinic, but was only found for the feeling informed subscale at the public clinic. The traditional DCS did not have adequate fit at T0 or at T2. The alternative DCS had adequate fit at T0 but was not identified at T2. Factor loadings indicated that two subscales, feeling informed and feeling clear about values, were not distinct factors. Conclusions. Our general PCS anxiety measure can be used in PCS decision aid studies. The alternative DCS may be appropriate for men eligible for PCS. Implications: More emphasis needs to be placed on the development of PCS anxiety items relating to testing procedures. We recommend that the two DCS versions be validated in other samples of men eligible for PCS and in other health care decisions that involve uncertainty. ^

Pathway Semantics: An Algebraic Data Driven Algorithm to Generate Hypotheses about Molecular Patterns Underlying Disease Progression

The overarching goal of the Pathway Semantics Algorithm (PSA) is to improve the in silico identification of clinically useful hypotheses about molecular patterns in disease progression. By framing biomedical questions within a variety of matrix representations, PSA has the flexibility to analyze combined quantitative and qualitative data over a wide range of stratifications. The resulting hypothetical answers can then move to in vitro and in vivo verification, research assay optimization, clinical validation, and commercialization. Herein PSA is shown to generate novel hypotheses about the significant biological pathways in two disease domains: shock / trauma and hemophilia A, and validated experimentally in the latter. The PSA matrix algebra approach identified differential molecular patterns in biological networks over time and outcome that would not be easily found through direct assays, literature or database searches. In this dissertation, Chapter 1 provides a broad overview of the background and motivation for the study, followed by Chapter 2 with a literature review of relevant computational methods. Chapters 3 and 4 describe PSA for node and edge analysis respectively, and apply the method to disease progression in shock / trauma. Chapter 5 demonstrates the application of PSA to hemophilia A and the validation with experimental results. The work is summarized in Chapter 6, followed by extensive references and an Appendix with additional material.

Caracterización del crecimiento de plantines de Brassica rapa var. pekinensis en contenedores.

Brassica rapa var. pekinensis, hakusay, es una hortaliza cuya producción puede iniciarse a partir de transplantes o siembra directa. El objetivo de este trabajo fue caracterizar el crecimiento de plantines de esta hortaliza en contenedores, previo al transplante, a través de variables morfogenéticas y de crecimiento. La biomasa total (PST), el peso seco aéreo (PSA) y el peso seco de raíces (PSR) mostraron una tendencia creciente, particular en cada caso. El área foliar (AF) se incrementó hasta estabilizarse días antes del transplante, momento en el cual se observó el efecto limitante del contenedor. Las variables morfogenéticas: razón de área foliar (RAF) y área foliar específica (AFE) descendieron hasta alcanzar valores casi constantes la semana previa al transplante. Las variables de crecimiento: índice de crecimiento relativo de la planta (ICRP) e índice de crecimiento relativo foliar (ICRF) descendieron, el primero con valores superiores hasta el día 31, hecho que se refleja a través de Gamma (Gf); la tasa absoluta de crecimiento (TAC) se incrementó hasta el día 31 del período y después cayó abruptamente; la tasa de asimilación neta (TAN) creció hasta el día 31 posterior a la siembra para luego caer al final del período. El productor podría decidir el momento del transplante como aquel definido por la detención de la expansión foliar.

Composition of rocks, minerals, water, and gases from uplift areas of the Central Atlantic

Original geological, geophysical, lithological, mineralogical data on uplifts of the Central Atlantic are given in the book based on materials of Cruise 1 of the R/V Akademik Nikolaj Strakhov. Geological and geophysical studies include description of the obtained material and analysis of structural and morphological elements of the ocean floor. Results of lithological, petrochemical and geochemical studies were extremely innovative and develop a conceptual model. The latter include studies of petrochemical evolution of tholeiitic alkaline plate volcanism, large-scale hydrothermal transformation of basement rocks - palygorskitization, phosphatization and ferromanganese mineralization. Showing imposition Superposition of hydrogenic alteration on hydrothermally altered rocks and its role in Cenozoic history of sedimentation is shown.

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.

Efficient algorithms for dynamic probabilistic safety assessment. An application to convex damage domain.

The design of nuclear power plant has to follow a number of regulations aimed at limiting the risks inherent in this type of installation. The goal is to prevent and to limit the consequences of any possible incident that might threaten the public or the environment. To verify that the safety requirements are met a safety assessment process is followed. Safety analysis is as key component of a safety assessment, which incorporates both probabilistic and deterministic approaches. The deterministic approach attempts to ensure that the various situations, and in particular accidents, that are considered to be plausible, have been taken into account, and that the monitoring systems and engineered safety and safeguard systems will be capable of ensuring the safety goals. On the other hand, probabilistic safety analysis tries to demonstrate that the safety requirements are met for potential accidents both within and beyond the design basis, thus identifying vulnerabilities not necessarily accessible through deterministic safety analysis alone. Probabilistic safety assessment (PSA) methodology is widely used in the nuclear industry and is especially effective in comprehensive assessment of the measures needed to prevent accidents with small probability but severe consequences. Still, the trend towards a risk informed regulation (RIR) demanded a more extended use of risk assessment techniques with a significant need to further extend PSA’s scope and quality. Here is where the theory of stimulated dynamics (TSD) intervenes, as it is the mathematical foundation of the integrated safety assessment (ISA) methodology developed by the CSN(Consejo de Seguridad Nuclear) branch of Modelling and Simulation (MOSI). Such methodology attempts to extend classical PSA including accident dynamic analysis, an assessment of the damage associated to the transients and a computation of the damage frequency. The application of this ISA methodology requires a computational framework called SCAIS (Simulation Code System for Integrated Safety Assessment). SCAIS provides accident dynamic analysis support through simulation of nuclear accident sequences and operating procedures. Furthermore, it includes probabilistic quantification of fault trees and sequences; and integration and statistic treatment of risk metrics. SCAIS comprehensively implies an intensive use of code coupling techniques to join typical thermal hydraulic analysis, severe accident and probability calculation codes. The integration of accident simulation in the risk assessment process and thus requiring the use of complex nuclear plant models is what makes it so powerful, yet at the cost of an enormous increase in complexity. As the complexity of the process is primarily focused on such accident simulation codes, the question of whether it is possible to reduce the number of required simulation arises, which will be the focus of the present work. This document presents the work done on the investigation of more efficient techniques applied to the process of risk assessment inside the mentioned ISA methodology. Therefore such techniques will have the primary goal of decreasing the number of simulation needed for an adequate estimation of the damage probability. As the methodology and tools are relatively recent, there is not much work done inside this line of investigation, making it a quite difficult but necessary task, and because of time limitations the scope of the work had to be reduced. Therefore, some assumptions were made to work in simplified scenarios best suited for an initial approximation to the problem. The following section tries to explain in detail the process followed to design and test the developed techniques. Then, the next section introduces the general concepts and formulae of the TSD theory which are at the core of the risk assessment process. Afterwards a description of the simulation framework requirements and design is given. Followed by an introduction to the developed techniques, giving full detail of its mathematical background and its procedures. Later, the test case used is described and result from the application of the techniques is shown. Finally the conclusions are presented and future lines of work are exposed.

Dinámica del uso del suelo y abastecimiento hídrico : implicaciones para la seguridad del agua en comunidades rurales en el Pacífico sur de Costa Rica.

En los países en desarrollo una de las prioridades que deben abordarse es la correcta gestión de los recursos hídricos que se ve amenazada por muchos factores, incluyendo el cambio rápido del uso del suelo, el cambio climático y, a menudo, la mala gestión. En este contexto se analiza la Cuenca Grande del rio Térraba en Costa Rica evaluando el cambio de uso del suelo desde 1986 a 2014 y, en particular, la evolución del conflicto de uso del suelo con respecto a su capacidad. Los cambios de uso del suelo pueden generar impactos relevantes tanto en la calidad del agua, como en los regíme-nes de caudales afectando a procesos esenciales, como la capacidad de retención de agua en el sue-lo, la escorrentía y la erosión. Para hacer frente a estos y otros impactos, el Ministerio de Salud de Costa Rica está desarrollando Planes de Seguridad del Agua (PSA) con una metodología propuesta por la OMS, promoviendo su adopción por todas las instituciones gestoras de agua para el consumo humano. A partir de las conclusiones del primer bloque, se elabora un PSA aplicado a un acueducto local que tiene problemas comunes a muchos de los sistemas de distribución rurales del Pacífico Sur de Costa Rica. En el marco de esta formulación se fortalece la institución gestora comunitaria (ASADA de Uvita y Bahía) y se evalúan las deficiencias que identificadas en este proceso, propo-niendo medidas correctivas. En el tercer bloque, profundizando en las medidas adoptadas para la seguridad del agua, se apoya a un segundo ente de distribución de agua (ASADA de Palmar Sur) con el diseño y proyecto de un sistema de desinfección para garantizar la seguridad contra la con-taminación bacteriana. Se evalúan diferentes alternativas y se selecciona la mejor tecnología adap-tada a las necesidades específicas de la ASADA mediante la evaluación de los recursos económicos y humanos potenciales del mismo ente.

Producción de Hidrógeno mediante energía solar térmica de alta temperatura.

Este proyecto versa sobre el estudio y diseño de una central termosolar de torre central, así como de su rendimiento y producción. La característica especial de esta planta es que tiene como fin suministrar calor a un proceso químico: cracking térmico del metano para la producción de hidrógeno. Debido a que el cracking térmico tiene lugar en el interior de un tanque de metal líquido (reactor químico) y al peso del mismo, resulta conveniente dejar el reactor a nivel de suelo. Así pues, los rayos solares tienen que descender desde la parte superior de la torre hasta el reactor. Los rayos solares se reflejan en los heliostatos para volver a ser reflejados en la parte superior de la torre, donde hay otro espejo que conduce los rayos solares hasta el reactor. La localización elegida para realizar el estudio ha sido Tabernas, cerca de la Plataforma Solar de Almería (PSA). De la estación SIAR (Sistema de Información Agroclimática para el Regadío del Ministerio de Agricultura, Alimentación y Medio Ambiente del Gobierno de España). de Tabernas precisamente se han obtenido los datos de radiación global, y mediante una correlación kd-kt, se ha obtenido la radiación directa. Para el estudio de la energía aportada por el campo solar, se ha elegido un campo norte de heliostatos, debido a que un campo circular además de tener menor rendimiento en este proyecto no sería factible. Los heliostatos considerados son cuadrados de 11 x 11 m. Estos heliostatos apuntan hacia el punto (0, 0, 100) m en el sistema de coordenadas absoluto considerado para el proyecto, que coincide con una altura a 100 m en el centro interior de la torre. Este punto es uno de los focos del elipsoide virtual, del cual forma parte el reflector (espejo situado en la parte superior) y cuyo otro foco se sitúa en la parte superior del receptor (reactor), cuyo fin es dirigir los rayos hacia el reactor, como se ha indicado. Una vez definido el proyecto, se lleva a cabo un dimensionado del campo solar, con el cual puede obtenerse un campo de heliostatos. Tras realizar la simulación, se obtienen datos instantáneos y medios. A modo de ejemplo, se incluye el rendimiento por bloqueos y sombras a las 9 h de la mañana en enero, donde puede observarse que la sombra de la torre tiene una alta importancia sobre los heliostatos situados en la zona oeste. Del mismo modo, se han obtenido los datos de incidencia de los rayos solares sobre el receptor, de forma que puede caracterizarse la incidencia del flujo térmico sobre el mismo mediante un mallado. A continuación se incluye una imagen del número de rayos solares que inciden sobre el receptor a las 12 h en junio. Las conclusiones que pueden extraerse del proyecto son:  El rendimiento anual de la planta es del 26 %.  La producción anual de hidrógeno sería de 163,7 t.  Para mayores potencias de plantas el rendimiento por sombras y bloqueos sería menor, al igual que el asociado al efecto coseno, a la dispersión y absorción atmosférica y al factor de interceptación.  Para reducir el efecto del tamaño en los dos primeros rendimientos mencionados en el párrafo anterior podría elevarse la altura de la torre.  Para aumentar el factor de interceptación podría estudiarse la colocación de espejos en el interior de la torre, de modo que reflejasen los rayos solares hasta el receptor.

Aplicación de la metodología ISA al análisis de la secuencia de pérdida del sistema de agua de refrigeración de componentes en una central nuclear con reactor de agua a presión

La metodología Integrated Safety Analysis (ISA), desarrollada en el área de Modelación y Simulación (MOSI) del Consejo de Seguridad Nuclear (CSN), es un método de Análisis Integrado de Seguridad que está siendo evaluado y analizado mediante diversas aplicaciones impulsadas por el CSN; el análisis integrado de seguridad, combina las técnicas evolucionadas de los análisis de seguridad al uso: deterministas y probabilistas. Se considera adecuado para sustentar la Regulación Informada por el Riesgo (RIR), actual enfoque dado a la seguridad nuclear y que está siendo desarrollado y aplicado en todo el mundo. En este contexto se enmarcan, los proyectos Safety Margin Action Plan (SMAP) y Safety Margin Assessment Application (SM2A), impulsados por el Comité para la Seguridad de las Instalaciones Nucleares (CSNI) de la Agencia de la Energía Nuclear (NEA) de la Organización para la Cooperación y el Desarrollo Económicos (OCDE) en el desarrollo del enfoque adecuado para el uso de las metodologías integradas en la evaluación del cambio en los márgenes de seguridad debidos a cambios en las condiciones de las centrales nucleares. El comité constituye un foro para el intercambio de información técnica y de colaboración entre las organizaciones miembro, que aportan sus propias ideas en investigación, desarrollo e ingeniería. La propuesta del CSN es la aplicación de la metodología ISA, especialmente adecuada para el análisis según el enfoque desarrollado en el proyecto SMAP que pretende obtener los valores best-estimate con incertidumbre de las variables de seguridad que son comparadas con los límites de seguridad, para obtener la frecuencia con la que éstos límites son superados. La ventaja que ofrece la ISA es que permite el análisis selectivo y discreto de los rangos de los parámetros inciertos que tienen mayor influencia en la superación de los límites de seguridad, o frecuencia de excedencia del límite, permitiendo así evaluar los cambios producidos por variaciones en el diseño u operación de la central que serían imperceptibles o complicados de cuantificar con otro tipo de metodologías. La ISA se engloba dentro de las metodologías de APS dinámico discreto que utilizan la generación de árboles de sucesos dinámicos (DET) y se basa en la Theory of Stimulated Dynamics (TSD), teoría de fiabilidad dinámica simplificada que permite la cuantificación del riesgo de cada una de las secuencias. Con la ISA se modelan y simulan todas las interacciones relevantes en una central: diseño, condiciones de operación, mantenimiento, actuaciones de los operadores, eventos estocásticos, etc. Por ello requiere la integración de códigos de: simulación termohidráulica y procedimientos de operación; delineación de árboles de sucesos; cuantificación de árboles de fallos y sucesos; tratamiento de incertidumbres e integración del riesgo. La tesis contiene la aplicación de la metodología ISA al análisis integrado del suceso iniciador de la pérdida del sistema de refrigeración de componentes (CCWS) que genera secuencias de pérdida de refrigerante del reactor a través de los sellos de las bombas principales del circuito de refrigerante del reactor (SLOCA). Se utiliza para probar el cambio en los márgenes, con respecto al límite de la máxima temperatura de pico de vaina (1477 K), que sería posible en virtud de un potencial aumento de potencia del 10 % en el reactor de agua a presión de la C.N. Zion. El trabajo realizado para la consecución de la tesis, fruto de la colaboración de la Escuela Técnica Superior de Ingenieros de Minas y Energía y la empresa de soluciones tecnológicas Ekergy Software S.L. (NFQ Solutions) con el área MOSI del CSN, ha sido la base para la contribución del CSN en el ejercicio SM2A. Este ejercicio ha sido utilizado como evaluación del desarrollo de algunas de las ideas, sugerencias, y los algoritmos detrás de la metodología ISA. Como resultado se ha obtenido un ligero aumento de la frecuencia de excedencia del daño (DEF) provocado por el aumento de potencia. Este resultado demuestra la viabilidad de la metodología ISA para obtener medidas de las variaciones en los márgenes de seguridad que han sido provocadas por modificaciones en la planta. También se ha mostrado que es especialmente adecuada en escenarios donde los eventos estocásticos o las actuaciones de recuperación o mitigación de los operadores pueden tener un papel relevante en el riesgo. Los resultados obtenidos no tienen validez más allá de la de mostrar la viabilidad de la metodología ISA. La central nuclear en la que se aplica el estudio está clausurada y la información relativa a sus análisis de seguridad es deficiente, por lo que han sido necesarias asunciones sin comprobación o aproximaciones basadas en estudios genéricos o de otras plantas. Se han establecido tres fases en el proceso de análisis: primero, obtención del árbol de sucesos dinámico de referencia; segundo, análisis de incertidumbres y obtención de los dominios de daño; y tercero, cuantificación del riesgo. Se han mostrado diversas aplicaciones de la metodología y ventajas que presenta frente al APS clásico. También se ha contribuido al desarrollo del prototipo de herramienta para la aplicación de la metodología ISA (SCAIS). ABSTRACT The Integrated Safety Analysis methodology (ISA), developed by the Consejo de Seguridad Nuclear (CSN), is being assessed in various applications encouraged by CSN. An Integrated Safety Analysis merges the evolved techniques of the usually applied safety analysis methodologies; deterministic and probabilistic. It is considered as a suitable tool for assessing risk in a Risk Informed Regulation framework, the approach under development that is being adopted on Nuclear Safety around the world. In this policy framework, the projects Safety Margin Action Plan (SMAP) and Safety Margin Assessment Application (SM2A), set up by the Committee on the Safety of Nuclear Installations (CSNI) of the Nuclear Energy Agency within the Organization for Economic Co-operation and Development (OECD), were aimed to obtain a methodology and its application for the integration of risk and safety margins in the assessment of the changes to the overall safety as a result of changes in the nuclear plant condition. The committee provides a forum for the exchange of technical information and cooperation among member organizations which contribute their respective approaches in research, development and engineering. The ISA methodology, proposed by CSN, specially fits with the SMAP approach that aims at obtaining Best Estimate Plus Uncertainty values of the safety variables to be compared with the safety limits. This makes it possible to obtain the exceedance frequencies of the safety limit. The ISA has the advantage over other methods of allowing the specific and discrete evaluation of the most influential uncertain parameters in the limit exceedance frequency. In this way the changes due to design or operation variation, imperceptibles or complicated to by quantified by other methods, are correctly evaluated. The ISA methodology is one of the discrete methodologies of the Dynamic PSA framework that uses the generation of dynamic event trees (DET). It is based on the Theory of Stimulated Dynamics (TSD), a simplified version of the theory of Probabilistic Dynamics that allows the risk quantification. The ISA models and simulates all the important interactions in a Nuclear Power Plant; design, operating conditions, maintenance, human actuations, stochastic events, etc. In order to that, it requires the integration of codes to obtain: Thermohydraulic and human actuations; Even trees delineation; Fault Trees and Event Trees quantification; Uncertainty analysis and risk assessment. This written dissertation narrates the application of the ISA methodology to the initiating event of the Loss of the Component Cooling System (CCWS) generating sequences of loss of reactor coolant through the seals of the reactor coolant pump (SLOCA). It is used to test the change in margins with respect to the maximum clad temperature limit (1477 K) that would be possible under a potential 10 % power up-rate effected in the pressurized water reactor of Zion NPP. The work done to achieve the thesis, fruit of the collaborative agreement of the School of Mining and Energy Engineering and the company of technological solutions Ekergy Software S.L. (NFQ Solutions) with de specialized modeling and simulation branch of the CSN, has been the basis for the contribution of the CSN in the exercise SM2A. This exercise has been used as an assessment of the development of some of the ideas, suggestions, and algorithms behind the ISA methodology. It has been obtained a slight increase in the Damage Exceedance Frequency (DEF) caused by the power up-rate. This result shows that ISA methodology allows quantifying the safety margin change when design modifications are performed in a NPP and is specially suitable for scenarios where stochastic events or human responses have an important role to prevent or mitigate the accidental consequences and the total risk. The results do not have any validity out of showing the viability of the methodology ISA. Zion NPP was retired and information of its safety analysis is scarce, so assumptions without verification or approximations based on generic studies have been required. Three phases are established in the analysis process: first, obtaining the reference dynamic event tree; second, uncertainty analysis and obtaining the damage domains; third, risk quantification. There have been shown various applications of the methodology and advantages over the classical PSA. It has also contributed to the development of the prototype tool for the implementation of the ISA methodology (SCAIS).

Análisis de secuencias de accidente de pérdida de refrigerante en centrales nucleares de agua a presión mediante la metodología ISA

El accidente de pérdida de refrigerante (LOCA) en un reactor nuclear es uno de los accidentes Base de Diseño más preocupantes y estudiados desde el origen del uso de la tecnología de fisión en la industria productora de energía. El LOCA ocupa, desde el punto de vista de los análisis de seguridad, un lugar de vanguardia tanto en el análisis determinista (DSA) como probabilista (PSA), cuya diferenciada perspectiva ha ido evolucionando notablemente en lo que al crédito a la actuación de las salvaguardias y las acciones del operador se refiere. En la presente tesis se aborda el análisis sistemático de de las secuencias de LOCA por pequeña y mediana rotura en diferentes lugares de un reactor nuclear de agua a presión (PWR) con fallo total de Inyección de Seguridad de Alta Presión (HPSI). Tal análisis ha sido desarrollado en base a la metodología de Análisis Integrado de Seguridad (ISA), desarrollado por el Consejo de Seguridad Nuclear (CSN) y consistente en la aplicación de métodos avanzados de simulación y PSA para la obtención de Dominios de Daño, que cuantifican topológicamente las probabilidades de éxito y daño en función de determinados parámetros inciertos. Para la elaboración de la presente tesis, se ha hecho uso del código termohidráulico TRACE v5.0 (patch 2), avalado por la NRC de los EEUU como código de planta para la simulación y análisis de secuencias en reactores de agua ligera (LWR). Los objetivos del trabajo son, principalmente: (1) el análisis exhaustivo de las secuencias de LOCA por pequeña-mediana rotura en diferentes lugares de un PWR de tres lazos de diseño Westinghouse (CN Almaraz), con fallo de HPSI, en función de parámetros de gran importancia para los transitorios, tales como el tamaño de rotura y el tiempo de retraso en la respuesta del operador; (2) la obtención y análisis de los Dominios de Daño para transitorios de LOCA en PWRs, de acuerdo con la metodología ISA; y (3) la revisión de algunos de los resultados genéricos de los análisis de seguridad para secuencias de LOCA en las mencionadas condiciones. Los resultados de la tesis abarcan tres áreas bien diferenciadas a lo largo del trabajo: (a) la fenomenología física de las secuencias objeto de estudio; (b) las conclusiones de los análisis de seguridad practicados a los transitorios de LOCA; y (c) la relevancia de las consecuencias de las acciones humanas por parte del grupo de operación. Estos resultados, a su vez, son de dos tipos fundamentales: (1) de respaldo del conocimiento previo sobre el tipo de secuencias analizado, incluido en la extensa bibliografía examinada; y (2) hallazgos en cada una de las tres áreas mencionadas, no referidos en la bibliografía. En resumidas cuentas, los resultados de la tesis avalan el uso de la metodología ISA como método de análisis alternativo y sistemático para secuencias accidentales en LWRs. ABSTRACT The loss of coolant accident (LOCA) in nuclear reactors is one of the most concerning and analized accidents from the beginning of the use of fission technology for electric power production. From the point of view of safety analyses, LOCA holds a forefront place in both Deterministic (DSA) and Probabilistic Safety Analysis (PSA), which have significantly evolved from their original state in both safeguard performance credibility and human actuation. This thesis addresses a systematic analysis of small and medium LOCA sequences, in different places of a nuclear Pressurized Water Reactor (PWR) and with total failure of High Pressure Safety Injection (HPSI). Such an analysis has been grounded on the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Regulatory Body (CSN). ISA involves the application of advanced methods of simulation and PSA for obtaining Damage Domains that topologically quantify the likelihood of success and damage regarding certain uncertain parameters.TRACE v5.0 (patch 2) code has been used as the thermalhydraulic simulation tool for the elaboration of this work. Nowadays, TRACE is supported by the US NRC as a plant code for the simulation and analysis of sequences in light water reactors (LWR). The main objectives of the work are the following ones: (1) the in-depth analysis of small and medium LOCA sequences in different places of a Westinghouse three-loop PWR (Almaraz NPP), with failed HPSI, regarding important parameters, such as break size or delay in operator response; (2) obtainment and analysis of Damage Domains related to LOCA transients in PWRs, according to ISA methodology; and (3) review some of the results of generic safety analyses for LOCA sequences in those conditions. The results of the thesis cover three separated areas: (a) the physical phenomenology of the sequences under study; (b) the conclusions of LOCA safety analyses; and (c) the importance of consequences of human actions by the operating crew. These results, in turn, are of two main types: (1) endorsement of previous knowledge about this kind of sequences, which is included in the literature; and (2) findings in each of the three aforementioned areas, not reported in the reviewed literature. In short, the results of this thesis support the use of ISA-like methodology as an alternative method for systematic analysis of LWR accidental sequences.

Cell–cell interaction in prostate gene regulation and cytodifferentiation

To examine the role of intercellular interaction on cell differentiation and gene expression in human prostate, we separated the two major epithelial cell populations and studied them in isolation and in combination with stromal cells. The epithelial cells were separated by flow cytometry using antibodies against differentially expressed cell-surface markers CD44 and CD57. Basal epithelial cells express CD44, and luminal epithelial cells express CD57. The CD57+ luminal cells are the terminally differentiated secretory cells of the gland that synthesize prostate-specific antigen (PSA). Expression of PSA is regulated by androgen, and PSA mRNA is one of the abundant messages in these cells. We show that PSA expression by the CD57+ cells is abolished after prostate tissue is dispersed by collagenase into single cells. Expression is restored when CD57+ cells are reconstituted with stromal cells. The CD44+ basal cells possess characteristics of stem cells and are the candidate progenitors of luminal cells. Differentiation, as reflected by PSA production, can be detected when CD44+ cells are cocultured with stromal cells. Our studies show that cell–cell interaction plays an important role in prostatic cytodifferentiation and the maintenance of the differentiated state.

From transforming growth factor-β signaling to androgen action: Identification of Smad3 as an androgen receptor coregulator in prostate cancer cells

Although transforming growth factor-β (TGF-β) has been identified to mainly inhibit cell growth, the correlation of elevated TGF-β with increasing serum prostate-specific antigen (PSA) levels in metastatic stages of prostate cancer has also been well documented. The molecular mechanism for these two contrasting effects of TGF-β, however, remains unclear. Here we report that Smad3, a downstream mediator of the TGF-β signaling pathway, functions as a coregulator to enhance androgen receptor (AR)-mediated transactivation. Compared with the wild-type AR, Smad3 acts as a strong coregulator in the presence of 1 nM 5α-dihydrotestosterone, 10 nM 17β-estradiol, or 1 μM hydroxyflutamide for the LNCaP mutant AR (mtAR T877A), found in many prostate tumor patients. We further showed that endogenous PSA expression in LNCaP cells can be induced by 5α-dihydrotestosterone, and the addition of the Smad3 further induces PSA expression. Together, our findings establish Smad3 as an important coregulator for the androgen-signaling pathway and provide a possible explanation for the positive role of TGF-β in androgen-promoted prostate cancer growth.

Application of principal component analysis in phase-shifting photoelasticity

Principal component analysis phase shifting (PCA) is a useful tool for fringe pattern demodulation in phase shifting interferometry. The PCA has no restrictions on background intensity or fringe modulation, and it is a self-calibrating phase sampling algorithm (PSA). Moreover, the technique is well suited for analyzing arbitrary sets of phase-shifted interferograms due to its low computational cost. In this work, we have adapted the standard phase shifting algorithm based on the PCA to the particular case of photoelastic fringe patterns. Compared with conventional PSAs used in photoelasticity, the PCA method does not need calibrated phase steps and, given that it can deal with an arbitrary number of images, it presents good noise rejection properties, even for complicated cases such as low order isochromatic photoelastic patterns. © 2016 Optical Society of America.