New mechanical samples positioning system for irradiations on a radial channel at nuclear research reactor in a full-power continuous operation

This paper describes a new mechanical samples positioning system that allows the safe placement and removal of biological samples for prolonged irradiation, in a nuclear reactor during full-power continuous operation. Also presented herein the materials of construction and operating principles. Additionally, this sample positioning system is compared with an existing pneumatic and automated transfer system, already available at the research reactors. The system consists of a mechanical arm with a claw, which can deliver the samples for irradiations without reactor shutdown. It was installed in the lEA-R1 research reactor at Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, Brazil, and for the past 5 years, the system has successfully operated and allowed the conducting of important experiments. As a result of its introduction, the facility has been in a position to positively respond to the increased demand in studies of biology, medicine, physics, engineering, detector/dosimeter calibrations, etc. It is one example of the appropriated technologies that save energy and resources. (C) 2010 Elsevier Ltd. All rights reserved.

New mechanical samples positioning system for irradiations on a radial channel at nuclear research reactor in a full-power continuous operation

FAPESP:97/5550

Radiocarbon impact in a nearby tree of a light-water VVER type nuclear power plant, Paks, Hungary

Tree-ring series were collected for radiocarbon analyses from the vicinity of Paks nuclear power plant (NPP) and a background area (Dunaföldvár) for a 10-yr period (2000–2009). Samples of holocellulose were prepared from the wood and converted to graphite for accelerator mass spectrometry (AMS) 14C measurement using the MICADAS at ETH Zürich. The 14C concentration data from these tree rings was compared to the background tree rings for each year. The global decreasing trend of atmospheric 14C activity concentration was observed in the annual tree rings both in the background area and in the area of the NPP. As an average of the past 10 yr, the excess 14C emitted by the pressurized-water reactor (PWR) NPP to the atmosphere shows only a slight systematic excess (~6‰) 14C in the annual rings. The highest 14C excess was 13‰ (in 2006); however, years with the same 14C level as the background were quite frequent in the tree-ring series.

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).

Dynamic evacuation analyses : independent assessments of evacuation times from the plume exposure pathway emergency planning zones of twelve nuclear power stations /

On cover: FEMA-REP-3.

Selected reactors of the Power Reactor Demonstration Program : a literature search /

U.S. Atomic Energy Commission Report No. TID-3556.

Selected reactors of the Power Reactor Demonstration Program : a literature search /

AEC Report No. TID-3556(REV. 1).

Selected reactors of the Power Reactor Demonstration Program : a literature search /

"Listing of reports dealing with eleven reactors of the Power Reactor Demonstration Program."

Evaluation of economic benefits from a continuing civilian nuclear power program /

SRIA-65, Reactor technology.

Draft environmental statement by the Directorate of Licensing, United States Atomic Energy Commission related to the proposed manufacture of floating nuclear power plants, Offshore Power Systems, Docket no. STN 50-437.

At head of title: v.2, NUREG-75/113; v.3, NUREG-0127.

Evaluation of radioiodine measurements at Pilgrim Nuclear Power Plant /

Manuscript completed September 1978, published October 1978.

The Fukushima nuclear power plant disaster and percetions of health risk communication : a case study

This case study was conducted to explore the perceptions of health risk messages sent by the Japanese Government following the Fukushima nuclear power plant disaster. The content of health risk messages from the Japanese Government and the Japanese national broadcaster (NHK) were analysed and semi-structured interviews were conducted with a sample of Tokyo residents. Initially, participants trusted these messages but as the crisis unfolded they became sceptical about the messages. Participants felt the messages did not communicate health risk information effectively because the messages were; not supported by evidence, inconsistent, delayed and changed over time. Despite widespread access to the internet, social media and mobile telephones, most participants relied on television news for information about the health risks. The Japanese Government urgently needs to re-build trust by engaging the community in the planning and development phases of health risk communication strategies.

Nuclear power and Japan's collective future: Understanding beliefs about how nuclear policies will affect society

Despite ongoing controversies regarding possible directions for the nuclear plants program throughout Japan since the Fukushima disaster, little has been researched about people's belief structure about future society and what may affect their attitudes toward different policy options. Beyond policy debates, the present study focused on how people see a future society according to the assumptions of different policy options. A total of 125 students at Japanese universities were asked to compare a future society with society today in which one of alternative policies was adopted (i.e., shutdown or expansion of nuclear reactors) in terms of characteristics of individuals and society in general. While perceived dangerousness of nuclear power predicted attitudes and behavioural intentions to make personal sacrifices for nuclear power policies, beliefs about the social consequences of the policies, especially on economic development and dysfunction, appeared to play stronger roles in predicting those measures. The importance of sociological dimensions in understanding how people perceive the future of society regarding alternative nuclear power policies, and the subtle discrepancies between attitudes and behavioural intentions, are discussed.

Assessing organizational culture in complex sociotechnical systems : Methodological evidence from studies in nuclear power plant maintenance organizations

Failures in industrial organizations dealing with hazardous technologies can have widespread consequences for the safety of the workers and the general population. Psychology can have a major role in contributing to the safe and reliable operation of these technologies. Most current models of safety management in complex sociotechnical systems such as nuclear power plant maintenance are either non-contextual or based on an overly-rational image of an organization. Thus, they fail to grasp either the actual requirements of the work or the socially-constructed nature of the work in question. The general aim of the present study is to develop and test a methodology for contextual assessment of organizational culture in complex sociotechnical systems. This is done by demonstrating the findings that the application of the emerging methodology produces in the domain of maintenance of a nuclear power plant (NPP). The concepts of organizational culture and organizational core task (OCT) are operationalized and tested in the case studies. We argue that when the complexity of the work, technology and social environment is increased, the significance of the most implicit features of organizational culture as a means of coordinating the work and achieving safety and effectiveness of the activities also increases. For this reason a cultural perspective could provide additional insight into the problem of safety management. The present study aims to determine; (1) the elements of the organizational culture in complex sociotechnical systems; (2) the demands the maintenance task sets for the organizational culture; (3) how the current organizational culture at the case organizations supports the perception and fulfilment of the demands of the maintenance work; (4) the similarities and differences between the maintenance cultures at the case organizations, and (5) the necessary assessment of the organizational culture in complex sociotechnical systems. Three in-depth case studies were carried out at the maintenance units of three Nordic NPPs. The case studies employed an iterative and multimethod research strategy. The following methods were used: interviews, CULTURE-survey, seminars, document analysis and group work. Both cultural analysis and task modelling were carried out. The results indicate that organizational culture in complex sociotechnical systems can be characterised according to three qualitatively different elements: structure, internal integration and conceptions. All three of these elements of culture as well as their interrelations have to be considered in organizational assessments or important aspects of the organizational dynamics will be overlooked. On the basis of OCT modelling, the maintenance core task was defined as balancing between three critical demands: anticipating the condition of the plant and conducting preventive maintenance accordingly, reacting to unexpected technical faults and monitoring and reflecting on the effects of maintenance actions and the condition of the plant. The results indicate that safety was highly valued at all three plants, and in that sense they all had strong safety cultures. In other respects the cultural features were quite different, and thus the culturally-accepted means of maintaining high safety also differed. The handicraft nature of maintenance work was emphasised as a source of identity at the NPPs. Overall, the importance of safety was taken for granted, but the cultural norms concerning the appropriate means to guarantee it were little reflected. A sense of control, personal responsibility and organizational changes emerged as challenging issues at all the plants. The study shows that in complex sociotechnical systems it is both necessary and possible to analyse the safety and effectiveness of the organizational culture. Safety in complex sociotechnical systems cannot be understood or managed without understanding the demands of the organizational core task and managing the dynamics between the three elements of the organizational culture.