Extraction of resins from WD-22 tank in Jose Cabrera Nuclear Power Plant

International Nuclear Information System (INIS)

Benavides, E.

1997-01-01

The Spent Resin Tank (WD-22) is located in the Auxiliary Building of Jose Cabrera Nuclear Plant (PWR 150 Mwe). This tank has a nominal capacity of 4 m 3 and is almost full of spent resins that has been stored from the late sixties to early eighties. As the lines are completely plugged with resins and due to the difficulties in the pipe lay-out, it has not been possible to transfer the resins to the cementation plant since that date. The plant decided, by an open bid quotation, to select the most suitable process to transfer the resins to the cementation plant avoiding the high doses existing in the tank cubicle and in a reasonable time schedule. The solution given in this paper contemplates that sometimes there is an imaginative answer to a problem that seems to be difficult to solve. (authors)

Performances and considerations of the study of environmental impact of the individualized temporary storage. Jose Cabrera NPP

International Nuclear Information System (INIS)

Cifuentes, N.; Hernandez Bravo, B.

2013-01-01

The dismantling phase of the Jose Cabrera nuclear power plant required a system able to manage the spent nuclear fuel produced during its activity: The individualized temporary storage. This project, which is pioneer in Spain, apart from being technically and securely friendly with the standards, should agree with sustainability standards. That is the reason why, while the technical project was being developed, Gas Natural Fenosa Engineering worked up an environmental impact assessment. This study, made by a multidisiciplimary team, analyzed all the interactions between the project and its environment, and allowed the inclusion of environmental measures that were required to reduce the negative effects and guarantee the environmental viability of the project. (Author)

Closing the sky. The total dismantling of the Jose Cabrera nuclear power plant demonstrates maturity in the nuclear sector

International Nuclear Information System (INIS)

Rodriguez, A.

2015-01-01

This article aims to put the situation of the decommissioning of nuclear power plants in the world into perspective as an already consolidated activity and with an important future of industrial activity. The decommissioning project that Enresa is currently performing in the old Jose Cabrera plant is being explained in detail, by providing data of the newest and most relevant technical aspects as well as the lessons learned to be reusable in other decommissioning projects. The previous background, the project planning, the activities performed and those still to be done as well as their timing are being explained in detail. (Author)

Dismantling and decommissioning of Jose Cabrera nuclear power plant

International Nuclear Information System (INIS)

Rodriguez, A.

2009-01-01

With the start of the dismantling works at the Jose Cabrera nuclear power plant now in sight, this is an appropriate moment to look back and consider recent history. The first time that the issue of nuclear power plant dismantling was dealt with was in 1975, at a conference in Paris entitled Nuclear Energy Maturity. Up until then the entire question had been one of design, construction and operation, but since that moment and it has been quite a while since that conference dismantling has begun to be seen as just another activity in the nuclear cycle, a final activity that will sooner or later affect all the facilities, an activity different from its predecessors and with the ultimate objective of restoring the sites for whatever use might be determined. During the 1960s and 1970s, the construction of nuclear power plants was widespread across the entire world. It was the baby boom of nuclear energy and now, forty or fifty years later, we are seeing the arrival of the end of the service lifetime of these plants and are faced with the corresponding general process of dismantling these installations. The dismantling of nuclear power plants has ceased to be an emerging issue and is now consolidated as a regular activity in the nuclear industry, albeit an activity that lacks adequate financing or specific regulation in certain countries. Fortunately this is not the case in Spain, since economic provisions have been planned and the regulatory framework developed. In view of the above, the dismantling of the nuclear power plants is an industrial activity involving specific technologies that implies new professional and business opportunities that should be absorbed and seized by society. In Spain the path followed in this direction has been a long one, as is underlined by the experiences of dismantling the Argos (Barcelona, 1998- 2004) and Arbi (Bilbao, 2002-2005) research reactors, the Andujar Uranium Mill (Jaen, 1991-1995), the Vandellos I nuclear power plant

Significant aspects of the external event analysis methodology of the Jose Cabrera NPP PSA

International Nuclear Information System (INIS)

Barquin Duena, A.; Martin Martinez, A.R.; Boneham, P.S.; Ortega Prieto, P.

1994-01-01

This paper describes the following advances in the methodology for Analysis of External Events in the PSA of the Jose Cabrera NPP: In the Fire Analysis, a version of the COMPBRN3 CODE, modified by Empresarios Agrupados according to the guidelines of Appendix D of the NUREG/CR-5088, has been used. Generic cases were modelled and general conclusions obtained, applicable to fire propagation in closed areas. The damage times obtained were appreciably lower than those obtained with the previous version of the code. The Flood Analysis methodology is based on the construction of event trees to represent flood propagation dependent on the condition of the communication paths between areas, and trees showing propagation stages as a function of affected areas and damaged mitigation equipment. To determine temporary evolution of the flood area level, the CAINZO-EA code has been developed, adapted to specific plant characteristics. In both the Fire and Flood Analyses a quantification methodology has been adopted, which consists of analysing the damages caused at each stage of growth or propagation and identifying, in the Internal Events models, the gates, basic events or headers to which safe failure (probability 1) due to damages is assigned. (Author)

Human factors inspection of current control room panel in Jose Cabrera NPP

International Nuclear Information System (INIS)

Almeida, P.; O'Hara, J.; Higgins, J.

2002-01-01

Within the process of renewal of Exploitation Permit of Jose Cabrera Nuclear Power Plant, UNION FENOSA GENERACIO, S. A. (UFG) has carried out an analysis and evaluation project regarding human factors implications of current control room panel arrangement. The project has been developed in two phases. In the first phase, leaded by EPRI and carried out by experts from SAIC, an independent review from a double viewpoint of human reliability and human factors was developed. In the second phase, a multidisciplinary team (composed by human factors, risk analysis, operation, engineering, training and instrumentation and controls experts) has developed a study on human factors implications of current panel arrangement, following the methodology pointed out in NUREG-0711. The project has been developed under the direction of Brookhaven National Laboratory (BNL), organisation that has authored the aforementioned methodology, with the participation of UFG and SOLUZIONA Ingenieria. For the development of the second study the following steps were taken: Firstly, the potential effects of panel arrangement on crew performance were identified its real evidence was analysed and the goals for the improvement of control room operation were established; following NUREG-0711. After this, several design alternatives that addressed these goals were identified and were analysed along three dimensions: human factors, risk analysis and economic costs. Finally the results of these evaluations were combined using a multi-attribute decision method to arrive at a recommended alternative as he best proposal to incorporate human factors criteria and good practices in the design of control room panels. (Author)

Latest feedback from a major reactor vessel dismantling project

International Nuclear Information System (INIS)

Boucau, J.; Segerud, P.; Sanchez, M.; Garcia, R.

2015-01-01

Westinghouse performed two large segmentation projects in 2010-2013 and then 2013-2015 at the Jose Cabrera nuclear power plant in Spain. The power plant is located in Almonacid de Zorita, 43 miles east of Madrid, Spain and was in operation between 1968 and 2006. This paper will describe the sequential steps required to prepare, segment, separate, and package the individual component segments using under water mechanical techniques. The paper will also include experiences and lessons learned that Westinghouse has collected from the activities performed during the reactor vessel and vessel internals segmentation projects. (authors)

Velocity of crack growing of Inconel-600, sensitized, contaminated with sulphur in PWR type reactors

International Nuclear Information System (INIS)

Castano, M. L.; Blazquez, F.; Gomez Briceno, D.; Lagares, A.

1998-01-01

The origin of the vessel head penetration cracking of Jose Cabrera NPP has been attributed to an IGA/SCC process in a highly sensitized Alloy 600 assisted by sulphur species, as both acid sulphates and reduced species originated by the thermal breakdown of the cationic resins present in the primary coolant. The thermal degradation of the cationic resins leads sulphonic acid group scission and sulphates. Under the operating conditions the reduction of sulphates to sulphides is produced. The sulphides formed from the reduction of sulphate can precipitate with metallic cations and be incorporated into the oxide layers of the materials, preferably into nickel alloys. Others components at Jose Cabrera NPP are fabricated from sensitized alloy 600, as bottom vessel penetrations. In order to determine the influence of sulphur incorporated to the oxide layers of bottom vessel penetration alloy 600, an experimental work has been performed to obtained crack growth rate data under PWR primary conditions on sensitized alloy 600. (Author) 5 refs

RAPD-PCR molecular analysis of the threatened Cabrera's vole ...

African Journals Online (AJOL)

Optimal management and conservation programs of the threatened Cabrera's vole require investigating potential molecular genetic markers in the genomic background, if the few remaining fragile populations are to ... The results described Cabrera's vole populations as a single genetic unit with slightly restricted gene flow.

Decommissioning Licensing Process of Nuclear Installations in Spain

International Nuclear Information System (INIS)

Correa Sainz, Cristina

2016-01-01

The Enresa experience related to the decommissioning of nuclear facilities includes the decommissioning of the Vandellos I and Jose Cabrera NPPs. The Vandellos I gas-graphite reactor was decommissioned in about five years (from 1998 to 2003) to what is known as level 2. In February 2010, the decommissioning of Jose Cabrera power plant has been initiated and it is scheduled to be finished by 2018. The decommissioning of a nuclear power plant is a complex administrative process, the procedure for changing from operation to decommissioning is established in the Spanish law. This paper summarizes the legal framework defining the strategies, the main activities and the basic roles of the various agents involved in the decommissioning of nuclear facilities in Spain. It also describes briefly the Licensing documents required to obtain the decommissioning authorization and the Enresa point of view, as licensee, on the licensing decommissioning process. (author)

New wireless system for fire protection (detection) during dismantling works in Jose Cabrera NPP (CNJC); Nuevo Sistema Inalámbrico de detección contraincendios para el desmantelamiento de a Central Nuclear José Cabrera

Energy Technology Data Exchange (ETDEWEB)

Núñez, F.J.; Gómez Rodríguez, C.A.

2016-07-01

José Cabrera Nuclear Power Plant is currently in an advanced stage in the process of dismantling and decommissioning. As part of that process, it is necessary to remove the maximum possible interference with operating systems, which are still needed to support the works ongoing or to ensure the safety. With that goal it has been installed a new wireless fire detection system that substitutes the old wired one. The coverage of this new system was initially focused on the containment building, establishing a first phase of analysis of signal coverage, and location of the different components in the building and a second phase of implementation (monitoring and testing) to demonstrate the reliability and robustness of the system in a building with such a complex geometry configuration. Currently the WIFI system is in operation, providing coverage to the containment building and other external facilities (laundry, control room, etc) that have been incorporated to support the process of dismantling.

Application of CSN Safety Guide on the content and criteria for the development of waste management plans at nuclear facilities; Aplicacion de la Guia de Seguridad del CSN sobre el contenido y criterio para la elaboracion de los planes de gestion de residuos en las instalaciones nucleares

Energy Technology Data Exchange (ETDEWEB)

Alvarez Alonso, M. A.; Simon Cirujano, M. I.

2011-07-01

The guide was developed with the joint participation and consensus of representatives of all Spanish companies who hold nuclear facilities and was implemented through a pilot project in the CN Jose Cabrera.

Un oficio del siglo XX: Guillermo Cabrera Infante en su laberinto fílmico

Directory of Open Access Journals (Sweden)

Javier Fernández Díaz

2016-12-01

Full Text Available El presente trabajo explora las relaciones entre literatura y cine en la obra de Guillermo Cabrera Infante. Para ello presenta dos bloques temáticos interrelacionados. En el primero se sugiere la lectura de Un oficio del siglo XX, un libro generalmente entendido como un conjunto de críticas cinematográficas, como un género híbrido, cercano al ensayo y a la autobiografía. El paratexto y las notas a pie de página del autor resultan clave. La segunda parte es una revisión de los tres guiones que Cabrera Infante escribió para cine a partir de la noción de texto fílmico.

The libertarian educational proposals of José Cabrera Díaz Las propuestas educativas libertarias de José Cabrera Díaz

Directory of Open Access Journals (Sweden)

Manuel FERRAZ LORENZO

2013-09-01

Full Text Available The purpose of this article is to explain the operational guidelines of the Canarian workers movement in the early Twentieth Century in connection with integral educational proposals. For this purpose, we have taken advantage of the articles published in the press by on of the leading exponents of the movement, a member of the anarchist school of thought, José Cabrera Diaz, whose curriculum vitae indicates his importance. He was a journalist, trades union leader and an outand- out defender of rationalist education and of the creation of people's schools, the purpose of which was to eradicate —by combating illiteracy and ignorance— the abusive quasi-feudal practices of the «caciques» that existed in the Canarian Archipiélago.El objetivo de este artículo consiste en exponer las líneas de trabajo del movimiento obrero en Canarias a comienzos de siglo, en lo concerniente a sus propuestas educativas de tipo integral. Para ello nos hemos servido de los trabajos publicados en prensa de uno de sus máximos representantes, inserto en la corriente de pensamiento anarquista: José Cabrera Díaz, cuyo curriculum lo dice todo. Periodista, líder sindical y, además, defensor a ultranza de la educación racionalista y de la creación de escuelas populares, destinadas a erradicar —desde la alfabetización y cultura— las prácticas caciquiles existentes en el Archipiélago.

Avance al estudio de las pinturas esquemáticas de las Peñas de Cabrera. Casabermeja, Málaga

Directory of Open Access Journals (Sweden)

Cecilio BARROSO RUIZ

2009-10-01

Full Text Available La estación de arte rupestre esquemático de las Peñas de Cabrera, se encuentra situada en el término municipal de Casabermeja, al este del mismo y muy cerca de la linde que separa a este pueblo con el de Colmenar. Su acceso es relativamente fácil, estando ubicada en el Cortijo de Cabrera (del cual toma el nombre; es un pequeño cerro de areniscas silíceas, muy duras y compactas, pertenecientes al Flysch de Colmenar.

Jose Vasconcelos, Philosopher-Librarian.

Science.gov (United States)

Christensen, Paul Martin

Jose Vasconcelos (1882-1959), a major figure of modern Mexican education, also was instrumental in the development of the field of librarianship in Mexico. This research paper presents Vasconcelos' philosophy of librarianship through a detailed study of his writings and a review of major secondary sources. Relying solely on the subjective…

Resistance Management for San Jose Scale (Hemiptera: Diaspididae).

Science.gov (United States)

Buzzetti, K; Chorbadjian, R A; Nauen, R

2015-12-01

The San Jose scale Diaspidiotus perniciosus Comstock is one of the most important pests of deciduous fruit trees. The major cause of recent outbreaks in apple orchards is thought to be the development of insecticide resistance, specifically organophosphates. The first report was given in North America, and now, in Chile. In the present study, San Jose scale populations collected from two central regions of Chile were checked for their susceptibility to different mode of action insecticides in order to establish alternatives to manage this pest. No evidence of cross resistance between organophosphates insecticides and acetamiprid, buprofezin, pyriproxyfen, spirotetramat, sulfoxaflor, or thiacloprid was found. Baselines of LC50-LC95 for different life stages of San Jose scale are given, as reference to future studies of resistance monitoring. The systemic activity of acetamiprid, spirotetramat, and thiacloprid was higher than the contact residue effect of these compounds. For sulfoxaflor, both values were similar. Program treatments including one or more of these compounds are compared in efficacy and impact on resistance ratio values. In order to preserve new insecticides as an important tool to control San Jose scale, resistance management programs should be implemented, considering insecticide mode of action classes alternated or mixed. © The Authors 2015. Published by Oxford University Press on behalf of Entomological Society of America. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Experience on Primary System Decommissioning in Jose Cabrera NPP

International Nuclear Information System (INIS)

Paloma Molleda; Leandro Sanchez; David Rodriguez

2015-01-01

Primary System Decommissioning belongs to DCP(Decommissioning and Closure Plan) works and its scope includes: Steam Generator, Pressurizer, Refrigerant Circuit Pump and Primary Circuit Piping. All these dismantling activities were carried out on site, including preliminary steps before their removal (SAS installations, pre decontaminations, cutting and segmentations, segregations, etc.) and delivery to media/low activity nuclear waste disposal site. There are many cutting techniques available in market (most of them proved with positive results) as well as there are many different approaches about how to manage radioactive wastes in decommissioning projects (containers or great components disposal, containers burial, re fusion, etc.). Both issues are linked and, before starting a new project, it might be positive and quite useful to compare and study previous dismantling experiences, especially the lesson learned chapter. Primary System cut with diamond saw has been a challenge target, not only due to the methodology innovation (since until nowadays, the common use of this technology was performed in cutting concrete walls) because it has a huge range of positive aspects that, in our opinion, are attractive (apart from its mentioned versatility, in terms of cutting on site and every type of material)

Experience on Primary System Decommissioning in Jose Cabrera NPP

Energy Technology Data Exchange (ETDEWEB)

Paloma Molleda; Leandro Sanchez; David Rodriguez [ENSA, Cantabria (Spain)

2015-10-15

Primary System Decommissioning belongs to DCP(Decommissioning and Closure Plan) works and its scope includes: Steam Generator, Pressurizer, Refrigerant Circuit Pump and Primary Circuit Piping. All these dismantling activities were carried out on site, including preliminary steps before their removal (SAS installations, pre decontaminations, cutting and segmentations, segregations, etc.) and delivery to media/low activity nuclear waste disposal site. There are many cutting techniques available in market (most of them proved with positive results) as well as there are many different approaches about how to manage radioactive wastes in decommissioning projects (containers or great components disposal, containers burial, re fusion, etc.). Both issues are linked and, before starting a new project, it might be positive and quite useful to compare and study previous dismantling experiences, especially the lesson learned chapter. Primary System cut with diamond saw has been a challenge target, not only due to the methodology innovation (since until nowadays, the common use of this technology was performed in cutting concrete walls) because it has a huge range of positive aspects that, in our opinion, are attractive (apart from its mentioned versatility, in terms of cutting on site and every type of material)

III Technical meeting of CAMP-ESPANA, Madrid 11 november 1999

International Nuclear Information System (INIS)

Veci Marrodan, I.; Lopez Jimenez, J.

2000-01-01

The program CAMP is leading by the Spanish Nuclear Safety Conseil (CSN). The program analyzes the validation and maintenance of thermohydraulic computer codes for the dynamic analysis of NNPPS. The Spanish participation are performed by; CSN, Almaraz, Asco, Cofrentes, Garona, Trillo, Vandellos and Jose Cabrera NNPPS, Universities and engineering companies: ENUSA, Tecnatom, CIEMAT, Empresarios Agrupados, UPV, UPC, UC, UPM

Supplementary study about the ATC generic project. Alternative storage for encapsulated fuel

International Nuclear Information System (INIS)

Canadas Martinez, I.

2010-01-01

The objective of this study is to present the alternatives for a complementary installation and facilitate the decision making on the best solution, keeping many determining factors in mind. Two are the alternatives studied: supplementary storage, similar to the Trillo Nuclear Power Plant, and outdoor storage, similar to the Asco and Jose Cabrera Nuclear Power Plants.

Predation on Alouatta guariba clamitans Cabrera (Primates, Atelidae by Leopardus pardalis (Linnaeus (Carnivora, Felidae Predação em Alouatta guariba clamitans Cabrera (Primates, Atelidae por Leopardus pordalis (Linnaeus (Carnivora, Felidae

Directory of Open Access Journals (Sweden)

João M. D. Miranda

2005-09-01

Full Text Available There are a few studies about predation on primates. Howler monkeys, being relatively large animals, were believed to be preyed on successfully only by medium to large-sized carnivores and large birds of prey. Our study took place at Chácara Payquerê, which is situated in the municipality of Balsa Nova, State of Paraná, Southern Brazil. Fingers and nails from Alouatta guariba clamitans Cabrera, 1940 were founded in two faecal samples from Leopardus pardalis (Linnaeus, 1758. With our documentation large howler monkey remains in faeces, the possibility of ocelot being a potential predator of all Neotropical primates should be taken into consideration.São poucos os estudos sobre a predação de primatas. Sendo os bugios animais relativamente grandes, acredita-se que somente médios e grandes carnívoros e grandes rapineiros possam lograr sucesso em sua predação. O estudo foi realizado na Chácara Payquerê, situada no município de Balsa Nova, Estado do Paraná, sul do Brasil. Dedos e unhas de Alouatta guariba clamitans Cabrera, 1940 foram encontradas em duas amostras fecais de Leopardus pardalis (Linnaeus, 1758 que foram coletadas. Com este registro de grande primata em fezes, a possibilidade de a jaguatirica ser um potencial predador de todos os primatas neotropicais pode ser levada em consideração.

Gochnatia polymorpha (Less.) Cabrera (Asteraceae) changes in leaf structure due to differences in light and edaphic conditions

OpenAIRE

Rossatto,Davi Rodrigo; Kolb,Rosana Marta

2010-01-01

Gochnatia polymorpha (Less.) Cabrera is a widespread Asteraceae species found in different physiognomies of cerrado (Neotropical savanna) and in forest formations of southeast Brazil. This study describes some leaf anatomy characteristics of this species and quantitatively evaluates them in relation to different environments, as well as under different light conditions. We found quantitative differences in all anatomical parameters analyzed. The results demonstrate that high leaf anatomy plas...

Amor, deseo y fantasía en El invitado de Carlos Arcos Cabrera

OpenAIRE

Anchante Arias, Jim

2015-01-01

El presente artículo es un análisis de la novela El invitado del escritor ecuatoriano Carlos Arcos Cabrera. Dicho análisis tiene como marco teórico el concepto lacaniano de fantasma, así como la noción de amor a partir de la propuesta ontológica de Alain Badiou. Se parte de la hipótesis de que en la mencionada novela, el matrimonio Felipe-Carmen configura una conflictiva relación que crea un discurso intrahistórico y otro histórico. Este último, teniendo como telón de fondo el conflicto de la...

Amor, deseo y fantasía en El invitado de Carlos Arcos Cabrera

OpenAIRE

Anchante Arias, Jim

2015-01-01

El presente artículo es un análisis de la novela El invitado del escritor ecuatoriano Carlos Arcos Cabrera. Dicho análisis tiene como marco teórico el concepto lacaniano de fantasma, así como la noción de amor a partir de la propuesta ontológica de Alain Badiou. Se parte de la hipótesis de que en la mencionada novela, el matrimonio Felipe- Carmen configura una conflictiva relación que crea un discurso intrahistórico y otro histórico. Este último, teniendo como telón de fondo el conflicto de l...

Inspection during operation of a nuclear power plant in Spain

International Nuclear Information System (INIS)

Gutierrez Bernal, R.

1977-01-01

The control and surveillance activities, as well as the operating data and results of the three nuclear power plants presently in operation: Jose Cabrera, Santa Maria de Garona and Vandellos, are summarized. The first two are light-water type, with different pressure and boiling characteristics and the third is of the gas-graphite type. The main aspects, from an inspection point of view, of the experience obtained in these three plants are analyzed. (author) [es

Characterization of aerosols in the Metropolitan Area of San Jose

International Nuclear Information System (INIS)

Mejias Perez, J.A.

1997-07-01

The objective of the present study, was to elaborate a profile of the contamination by private matter and to characterize the aerosols collected in the Metropolitan Area of San Jose (Costa Rica). For that, a campaign of sampling was carried out in three points of the city of San Jose, differentiated by there degree of activity: Center of San Jose (Central Station of Firemen), San Isidro of Coronado -Canton of Vasquez of Coronado- (Municipality) and Escazu (Municipality). Such campaign was carried out from April 4 to July 4, 1996 (transition summer-winter), and in two periods of time of 8 hours: 8 a.m. to 4 p.m. and of 8 p.m. to 4 a.m. The aerosols were collected utilizing Gent Pm-10 samplers, in filters of polycarbonate of 0,4 μm and 8 μm in cascade, with a flow average of 15 L/min., and it determined the composition average of the present aerosols. The concentration of the majority of the anions were obtained by means of ionic chromatography of high resolution, and the main cations by spectrophotometry of atomic absorption with electro thermic atomization. The space-temporary variations of the concentrations were evaluated and their correlation with the meteorologic variable. (S. Grainger) [es

Geologic mapping around Mahoma mining. San Jose mining company;Carta geologica de los alrededores de Mina Mahoma, Compania Minera San Jose

Energy Technology Data Exchange (ETDEWEB)

Techera, J; Arrighetii, R

1993-07-01

This study has as main objective carry out a geological mapping as well as the structural analysis , in 1.5.000 scale in the zone where the gold benefit plant of San Jose mining company is settled (Mahoma Mining). From this study has been marked many drillings.

Gochnatia polymorpha (Less. Cabrera (Asteraceae changes in leaf structure due to differences in light and edaphic conditions Gochnatia polymorpha (Less. Cabrera modifica sua estrutura foliar devido a diferenças na luminosidade e condições edáficas

Directory of Open Access Journals (Sweden)

Davi Rodrigo Rossatto

2010-09-01

Full Text Available Gochnatia polymorpha (Less. Cabrera is a widespread Asteraceae species found in different physiognomies of cerrado (Neotropical savanna and in forest formations of southeast Brazil. This study describes some leaf anatomy characteristics of this species and quantitatively evaluates them in relation to different environments, as well as under different light conditions. We found quantitative differences in all anatomical parameters analyzed. The results demonstrate that high leaf anatomy plasticity is an adaptive advantage that allows this species to occur in diverse cerrado conditions.Gochnatia polymorpha (Less. Cabrera é uma espécie de Asteraceae com ampla distribuição no bioma cerrado, sendo encontrada em diversas fisionomias florestais da região sudeste do Brasil. O presente estudo descreve alguns caracteres anatômicos foliares dessa espécie e os analisa quantitativamente em função de sua ocorrência nas formações florestais e também das diferenças de luminosidade. Foram encontradas diferenças quantitativas em todos os parâmetros anatômicos analisados. Os resultados demonstram que a alta plasticidade anatômica foliar nesta espécie pode ser considerada como uma vantagem adaptativa que a permite ocorrer em diversos ambientes do cerrado.

San Jose Accord: energy aid or petroleum-marketing strategy

Energy Technology Data Exchange (ETDEWEB)

1982-09-30

The San Jose Accord was signed in San Jose, Costa Rica on August 3, 1980 by the Presidents of Venezuela and Mexico, whereby the two countries mutually committed to supply the net imported domestic oil consumption of several Central American and Caribbean countries. Countries initially participating in the program are: Barbados, Costa Rica, Dominican Republic, El Salvador, Guatemala, Honduras, Jamaica, Nicaragua, and Panama. Seven eastern Caribbean countries were to meet on October 7 to petition for inclusion in the Accord, namely: Antigua, St. Kitt/Nevis, Montserrat, Dominica, St. Lucia, St. Vincent, and Grenada. The official language of the Accord is presented, and the operative status of the Accord two years after signing is discussed. Specific briefs about some of the individual countries in the Accord are included. The fuel price/tax series for the Western Hemisphere countries is updated.

La Biblioteca Latino Americana: User Survey (San Jose Public Library). Studies in Librarianship No. 2.

Science.gov (United States)

Johnstone, James C.; And Others

To assist a neighborhood committee in applying for federal funding of a bilingual/bicultural library with a distinct Latin American emphasis, a student research group from San Jose State University designed and administered a bilingual questionnaire to a stratified sample of 400 households in the Gardner District of San Jose, California. The…

Arthropod gut symbionts from the Balearic Islands: Majorca and Cabrera. Diversity and biogeography

Directory of Open Access Journals (Sweden)

Guàrdia Valle, Laia

2009-12-01

Full Text Available This study includes a catalogue with all the current data concerning the presence of trichomycetes (sensu lato in Majorca and Cabrera, as well as information on the biology, ecology and biogeographic implications of the insularity for each taxon of these arthropod-gut symbionts. Of the 13 species here reported, 10 are new for the Balearic Islands, including 4 Mesomycetozoan, of which 3 Eccrinales (Astreptonema gammari, Eccrinidus flexilis, Parataeniella dilatata, 1 Amoebidiales (Paramoebidium curvum and 6 kixckellomycotina Harpellales (Genistellospora homothallica, Harpella melusinae, Smittium culisetae, S. simulii, Stachylina grandispora and St. nana; the additional 3 were previously reported elsewhere: Asellaria ligiae (Aslleariales, Legeriomyces rarus and Stipella vigilans (Harpellales, but are here included as indissoluble part of the present Balearic catalogue. All taxa are commented, illustrated and their biogeographic implications are discussed.

El presente estudio incluye una recopilación de todos los datos concernientes al conocimiento de los tricomicetos (sensu lato en las islas Baleares de Mallorca y Cabrera, incluyendo un catálogo de especies y notas sobre la biología, ecología e implicaciones biogeográficas de su insularidad. De las 13 especies citadas, 10 son nuevas para las Baleares, incluyendo 4 Mesomycetozoos, de los cuales 3 Eccrinales (Astreptonema gammari, Eccrinidus flexilis, Parataeniella dilatata, 1 Amoebidiales (Paramoebidium curvum y 6 Harpellales (kixckellomycotina (Genistellospora homothallica, Harpella melusinae, Smittium culisetae, S. simulii, Stachylina grandispora y St. nana; aunque las 3 especies restantes: Asellaria ligiae (Aslleariales, Legeriomyces rarus y Stipella vigilans (Harpellales fueron citadas anteriormente, se incluyen aquí brevemente como parte del catálogo.

IMPROVING CONTROL ROOM DESIGN AND OPERATIONS BASED ON HUMAN FACTORS ANALYSES OR HOW MUCH HUMAN FACTORS UPGRADE IS ENOUGH ?

Energy Technology Data Exchange (ETDEWEB)

HIGGINS,J.C.; OHARA,J.M.; ALMEIDA,P.

2002-09-19

THE JOSE CABRERA NUCLEAR POWER PLANT IS A ONE LOOP WESTINGHOUSE PRESSURIZED WATER REACTOR. IN THE CONTROL ROOM, THE DISPLAYS AND CONTROLS USED BY OPERATORS FOR THE EMERGENCY OPERATING PROCEDURES ARE DISTRIBUTED ON FRONT AND BACK PANELS. THIS CONFIGURATION CONTRIBUTED TO RISK IN THE PROBABILISTIC SAFETY ASSESSMENT WHERE IMPORTANT OPERATOR ACTIONS ARE REQUIRED. THIS STUDY WAS UNDERTAKEN TO EVALUATE THE IMPACT OF THE DESIGN ON CREW PERFORMANCE AND PLANT SAFETY AND TO DEVELOP DESIGN IMPROVEMENTS.FIVE POTENTIAL EFFECTS WERE IDENTIFIED. THEN NUREG-0711 [1], PROGRAMMATIC, HUMAN FACTORS, ANALYSES WERE CONDUCTED TO SYSTEMATICALLY EVALUATE THE CR-LA YOUT TO DETERMINE IF THERE WAS EVIDENCE OF THE POTENTIAL EFFECTS. THESE ANALYSES INCLUDED OPERATING EXPERIENCE REVIEW, PSA REVIEW, TASK ANALYSES, AND WALKTHROUGH SIMULATIONS. BASED ON THE RESULTS OF THESE ANALYSES, A VARIETY OF CONTROL ROOM MODIFICATIONS WERE IDENTIFIED. FROM THE ALTERNATIVES, A SELECTION WAS MADE THAT PROVIDED A REASONABLEBALANCE BE TWEEN PERFORMANCE, RISK AND ECONOMICS, AND MODIFICATIONS WERE MADE TO THE PLANT.

Geologic mapping around Mahoma mining. San Jose mining company

International Nuclear Information System (INIS)

Techera, J.; Arrighetii, R.

1993-01-01

This study has as main objective carry out a geological mapping as well as the structural analysis , in 1.5.000 scale in the zone where the gold benefit plant of San Jose mining company is settled (Mahoma Mining). From this study has been marked many drillings.

Parcours croisés de Dominique Cabrera, cinéaste, et de ses proches collaborateurs. Intersecting careers: Dominique Cabrera, film director, and her close collaborators.

Directory of Open Access Journals (Sweden)

Adeline Lamberbourg

2010-07-01

Full Text Available Le métier de réalisateur de cinéma a été peu étudié en sciences humaines et sociales. Les études cinématographiques et la presse spécialisée s’intéressent beaucoup moins aux parcours des collaborateurs de l’équipe technique qu’aux comédiens et aux projets des auteurs que l’on pourrait croire retranchés dans leur intériorité créatrice. Pour échapper à ce travers biographique, le sociologue peut rendre compte de l’itinéraire d’un réalisateur en examinant ses liens avec ses collaborateurs. On prend ici l’exemple des liens de travail récurrents qui unissent Dominique Cabrera, réalisatrice de cinéma, à ses proches collaborateurs, en montrant de quelle manière des liens de travail peuvent s’approfondir en dépit d’un contexte sectoriel où l’organisation par projet de l’activité cinématographique favorise a priori la fugacité des relations de travail. On postule ainsi que le contexte pertinent pour éclairer les contours de cet itinéraire de carrière est le réseau des relations socioprofessionnelles au sein duquel naissent les projets de films et qui contribuent à ancrer ce parcours dans le paysage professionnel des pratiques cinématographiques.In France, social sciences and humanities have paid little attention to the profession of film director. Film studies and the press are much less interested in the professional careers of technical teams in comparison with the lives of actors and authors of projects who seem to be cut into their inner creativity. To escape the biographical pitfall, sociologists may account for a director’s itinerary by analyzing his/her ties with colleagues. Here, we examine the recurrent work relations between film director Dominique Cabrera and her close colleagues, to underline the strong relationships which paradoxically unite both the director and her staff in a project-based industry that is rather conducive to short-lived work relations. We postulate that the

NPP Evaluation, backfitting and life extension. An engineering viewpoint

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Lopez, A [Empresarios Agrupados, A.I.E., Madrid (Spain)

1993-12-15

During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

NPP Evaluation, backfitting and life extension. An engineering viewpoint

International Nuclear Information System (INIS)

Gonzalez Lopez, A.

1993-01-01

During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

Tsenõ rastut, poka ljudi pokupajut / Marje Josing ; interv. Eteri Kekelidze

Index Scriptorium Estoniae

Josing, Marje, 1959-

2007-01-01

Konjunktuuriinstituudi direktor Marje Josing selgitab inflatsiooni ja tarbijahinnaindeksi mõistete erinevusi, palgatõusu ja inflatsiooni kasvu põhjusi, Eesti majanduse üldisi arenguid ja annab lähiaastate prognoosi

San Jose, California: Evaluating Local Solar Energy Generation Potential (City Energy: From Data to Decisions)

Energy Technology Data Exchange (ETDEWEB)

Office of Strategic Programs, Strategic Priorities and Impact Analysis Team

2017-09-29

This fact sheet "San Jose, California: Evaluating Local Solar Energy Generation Potential" explains how the City of San Jose used data from the U.S. Department of Energy's Cities Leading through Energy Analysis and Planning (Cities-LEAP) and the State and Local Energy Data (SLED) programs to inform its city energy planning. It is one of ten fact sheets in the "City Energy: From Data to Decisions" series.

75 FR 29722 - Foreign-Trade Zone 18-San Jose, CA; Application for Subzone; Lam Research Corporation (Wafer...

Science.gov (United States)

2010-05-27

... DEPARTMENT OF COMMERCE Foreign-Trade Zones Board [Docket 36-2010] Foreign-Trade Zone 18--San Jose, CA; Application for Subzone; Lam Research Corporation (Wafer Fabrication Equipment Manufacturing); Fremont, Newark, and Livermore, CA An application has been submitted to the Foreign-Trade Zones Board (the Board) by the City of San Jose, grantee of...

Kepentingan Pemerintah Uruguay Melegalisasi Ganja Pada Masa Pemerintahan Jose Alberto Mujica Cordano Tahun 2010-2015

OpenAIRE

Nainggolan, Pebrianto; Yelta, Den

2015-01-01

This study explain about Uruguay government interests to legalization marijuana in the reign of Jose Alberto Mujica Cordano in 2013. Uruguay is a first country who legalize cultivation, distribution, and consumtion of marijuana in the world. Marijuana is plant which producing a fiber, but better known as narcotic substance in the seeds that can make the user feeling pleasure without cause. This regulation was first proposed by Jose Alberto Mujica Cordano and approved by parlament and senate. ...

Purificação do flavonóide trans-tilirosídeo do extrato metanólico das folhas de Gochnatia barrosii Cabrera (asteraceae e avaliação da sua atividade nematicida Purification of the flavonoid trans-tiliroside from the methanolic extract of Gochnatia barrosii Cabrera (Asteraceae leaves and evaluation of the nematicidal activity

Directory of Open Access Journals (Sweden)

Helvécio Martins dos Santos Júnior

2010-10-01

Full Text Available O fracionamento do extrato metanólico das folhas de Gochnatia barrosii Cabrera (Asteraceae resultou no isolamento do flavonol glicosídico trans-tilirosídeo [kaempferol 3-O- -D-(6''-O-E-p-cumaroil-glicopiranosídeo], que nunca tinha sido identificado na referida espécie vegetal. Em teste realizado in vitro, observou-se que tal substância a 500 μg/mL, não tem efeito sobre a mortalidade de juvenis do segundo estágio do nematóide Meloidogyne exigua Goeldi.Fractionation of the methanolic extract from Gochnatia barrosii Cabrera (Asteraceae leaves resulted in the isolation of the flavonol glycoside trans-tiliroside [kaempferol 3-O- -D-(6''-O-E-p-coumaroyl-glucopyranoside], which had never been found in such plant species. Such substance at 500 μg/mL caused no in vitro effect on the mortality of second-stage juveniles of the nematode Meloidogyne exigua Goeldi.

Jose f Regional Underground Research Centre: a new and attractive location for interdisciplinary teaching, research and training in the field of nuclear engineering

International Nuclear Information System (INIS)

Pacovsky, J.; Vasicek, R.

2010-10-01

The Jose f Gallery, located in the central Bohemia region of the Czech Republic (not far from the capital, Prague), was first excavated in 1981 as an exploration complex for the potential mining of gold. In 2007, the gallery was substantially reconstructed to house the Jose f Underground Educational Facility (Jose f UEF), which subsequently became an autonomous workplace under the direction of the Czech Technical University in Prague. At the beginning of 2010, the UEF was renamed the Jose f Regional Underground Research Centre (Jose f URC) which, along with the extensive underground complex, features modern above-ground facilities. One of the most important roles of this research centre is to provide practical -in situ- instruction in the fields of geotechnical engineering, geology, geochemistry, radiochemistry and radioecology. The training of future experts in this authentic underground setting involves the participation of several other Czech universities and numerous experienced specialists from outside the academic sphere. The IAEA has recently added the Jose f URC to its prestigious list of international training centres involved in the Training in and demonstration of waste disposal technologies in underground research facilities - A network of centres of excellence project. In addition to teaching and training, the Jose f URC is heavily involved in a wide range of research and development activities. The Jose f URC underground facilities are currently being used for research purposes as part of several European Union International experimental projects addressing various issues related to deep repository radioactive waste disposal (TIMODAZ - FP6, Forge - FP7, PETRUS II - FP7) as well as for hosting domestic projects supported by the Czech Ministry of Industry and Trade and the Czech Science Foundation. The Jose f URC is also working in close cooperation with the private construction sector providing practical training in underground construction

Jose f Regional Underground Research Centre: a new and attractive location for interdisciplinary teaching, research and training in the field of nuclear engineering

Energy Technology Data Exchange (ETDEWEB)

Pacovsky, J.; Vasicek, R., E-mail: Pacovsky@fsv.cvut.c [Czech Technical University in Prague, Faculty of Civil Engineering, Centre of Experimental Geotechnics, Thakurova 7, 166-29 Prague 6 (Czech Republic)

2010-10-15

The Jose f Gallery, located in the central Bohemia region of the Czech Republic (not far from the capital, Prague), was first excavated in 1981 as an exploration complex for the potential mining of gold. In 2007, the gallery was substantially reconstructed to house the Jose f Underground Educational Facility (Jose f UEF), which subsequently became an autonomous workplace under the direction of the Czech Technical University in Prague. At the beginning of 2010, the UEF was renamed the Jose f Regional Underground Research Centre (Jose f URC) which, along with the extensive underground complex, features modern above-ground facilities. One of the most important roles of this research centre is to provide practical -in situ- instruction in the fields of geotechnical engineering, geology, geochemistry, radiochemistry and radioecology. The training of future experts in this authentic underground setting involves the participation of several other Czech universities and numerous experienced specialists from outside the academic sphere. The IAEA has recently added the Jose f URC to its prestigious list of international training centres involved in the Training in and demonstration of waste disposal technologies in underground research facilities - A network of centres of excellence project. In addition to teaching and training, the Jose f URC is heavily involved in a wide range of research and development activities. The Jose f URC underground facilities are currently being used for research purposes as part of several European Union International experimental projects addressing various issues related to deep repository radioactive waste disposal (TIMODAZ - FP6, Forge - FP7, PETRUS II - FP7) as well as for hosting domestic projects supported by the Czech Ministry of Industry and Trade and the Czech Science Foundation. The Jose f URC is also working in close cooperation with the private construction sector providing practical training in underground construction

ETNOGRAFÍA ARGENTINA: LA CÁTEDRA LIBRE DE MONSEÑOR PABLO CABRERA (1925 / Argentine Ethnography: Monsignor Pablo Cabrera’s free lecture (1925

Directory of Open Access Journals (Sweden)

Mariela Eleonora Zabala

2010-12-01

Full Text Available La Reforma Universitaria de la Universidad Nacional de Córdoba (UNC en 1918, posibilitó la inserción académica de la antropología en el área central argentina, mediante dos iniciativas de los estudiantes. Una fueron los cursos y docencia libre de carácter extracurricular, para alumnos de todas las carreras. Otra fueron los concursos docentes, destinados a excluir de la autoridad, la administración y el cuerpo docente a los católicos y conservadores, y beneficiar a los profesores “más capaces y dignos” en nombre de “la idoneidad”, “la justicia” y “la libertad”.Este artículo examina una derivación directa de estas iniciativas que parece contradecir sus propósitos: la emergencia del primer curso libre de Etnografía Argentina en la UNC en 1925. Dicho espacio curricular fue propuesto y estuvo a cargo de Monseñor Pablo Cabrera (1857-1936, sacerdote de la diócesis de Córdoba, miembro de la Academia Nacional de Ciencias, director del Museo Histórico Provincial, y prolífero autor de libros y artículos publicados en la Revista de la Universidad y el diario Los Principios. ¿Cómo fue posible que un intelectual de la Iglesia encabezara semejante innovación en un contexto anticlerical? ¿Cómo incidió su adscripción confesional en una institución laica, y cómo incidió su obra universitaria en su vertiente clerical? ¿Cómo convivieron estas dos líneas en una misma persona, precursor de la antropología cordobesa? Palabras Claves: Reforma universitaria; enseñanza; Etnología argentina; Monseñor Pablo Cabrera. AbstractThe groundbreaking Reform that took place at the National University of Córdoba (Universidad Nacional de Córdoba, UNC in 1918, paved the way to the academic start of anthropology in Central Argentina. This new presence was fostered by two initiatives of the students’ sector: extra-curriculum or “free” courses, and public contests to professors’ positions. The latter was meant to exclude

Spain investigates PLEX options

International Nuclear Information System (INIS)

Hevia, F.

1990-01-01

Spain's nuclear generation capacity will be reduced by some 6000MWe by the year 2015 if decommissioning of the units currently in operation takes place at the end of their 40-year design life. Bearing this in mind, in 1988 the Santa Maria de Garona BWR and the Jose Cabrera PWR were chosen by their respective owners as reference units for plant life extension (PLEX) activities. These plants are the oldest of their types operating in Spain and PLEX programmes were already under way. (author)

Best practices for preparing vessel internals segmentation projects

International Nuclear Information System (INIS)

Boucau, Joseph; Segerud, Per; Sanchez, Moises

2016-01-01

Westinghouse has been involved in reactor internals segmentation activities in the U.S. and Europe for 30 years. Westinghouse completed in 2015 the segmentation of the reactor vessel and reactor vessel internals at the Jose Cabrera nuclear power plant in Spain and a similar project is on-going at Chooz A in France. For all reactor dismantling projects, it is essential that all activities are thoroughly planned and discussed up-front together with the customer. Detailed planning is crucial for achieving a successful project. One key activity in the preparation phase is the 'Segmentation and Packaging Plan' that documents the sequential steps required to segment, separate, and package each individual component, based on an activation analysis and component characterization study. Detailed procedures and specialized rigging equipment have to be developed to provide safeguards for preventing certain identified risks. The preparatory work can include some plant civil structure modifications for making the segmentation work easier and safer. Some original plant equipment is sometimes not suitable enough and need to be replaced. Before going to the site, testing and qualification are performed on full scale mock-ups in a specially designed pool for segmentation purposes. The mockup testing is an important step in order to verify the function of the equipment and minimize risk on site. This paper is describing the typical activities needed for preparing the reactor internals segmentation activities using under water mechanical cutting techniques. It provides experiences and lessons learned that Westinghouse has collected from its recent projects and that will be applied for the new awarded projects. (authors)

Läänemere strateegia arvestab kodanike ettepanekutega / Jose Palma Andres ; interv. Erkki Bahovski

Index Scriptorium Estoniae

Palma Andres, Jose

2008-01-01

Euroopa Komisjoni regionaalpoliitika peadirektoraadi direktor Jose Palma Andres vastab küsimustele, mis puudutavad avalikku debatti Läänemere strateegia üle ning Venemaa partnerlust Läänemere strateegias. Lisa: [Läänemere strateegia

On the Connection between Kinetic Monte Carlo and the Burton-Cabrera-Frank Theory

Science.gov (United States)

Patrone, Paul; Margetis, Dionisios; Einstein, T. L.

2013-03-01

In the many years since it was first proposed, the Burton- Cabrera-Frank (BCF) model of step-flow has been experimentally established as one of the cornerstones of surface physics. However, many questions remain regarding the underlying physical processes and theoretical assumptions that give rise to the BCF theory. In this work, we formally derive the BCF theory from an atomistic, kinetic Monte Carlo model of the surface in 1 +1 dimensions with one step. Our analysis (i) shows how the BCF theory describes a surface with a low density of adsorbed atoms, and (ii) establishes a set of near-equilibrium conditions ensuring that the theory remains valid for all times. Support for PP was provided by the NIST-ARRA Fellowship Award No. 70NANB10H026 through UMD. Support for TLE and PP was also provided by the CMTC at UMD, with ancillary support from the UMD MRSEC. Support for DM was provided by NSF DMS0847587 at UMD.

El hipotexto figurativo en “Vista del amanecer en el trópico” de Guillermo Cabrera Infante

Directory of Open Access Journals (Sweden)

Alessandra Ghezzani

2015-07-01

Full Text Available Vista del amanecer en el trópico narrates Cuban history through its most important steps. The book is composed by 101 fragments or “viñetas”; some of them are descriptions of images (lithographies, drawings, paintings, photographs related to characters or events of that history. This article presents the results of a research aimed at identifying those “absent images” (a subject that critics have up to now ignored, discovering the sources exploited by the author Cabrera Infante, highlighting the discourse strategies underlying their verbal representation and, finally, assessing how the image selection and their representation contribute to shape the vision of cuban history that comes to light from the book.

Metallothionein and heavy metals in daphnia pulex from Jose Antonio Alzate reservoir

International Nuclear Information System (INIS)

Avila Perez, P.; Zarazua Ortega, G.; Barcelo Quintal, D.; Rosas, I.; Diazdelgado, C.

2001-01-01

Water and specimens of the freshwater cladoceran Dhapnia pulex were collected at 4 different sites located in an area influenced by industrial, agricultural and urban activities in the Jose Antonio Alzate Reservoir in two different seasons. The Jose Antonio Alzate Reservoir fed by the Lerma river is the first significant water reservoir downstream of the main industrial areas in the State of Mexico. There are about 2,500 industrial discharges between the river source and the Alzate Reservoir which makes the Lerma river and the Jose Antonio Alzate Reservoir the most contaminated water bodies in the State of Mexico. The Monitoring National Network recognises these waters as highly contaminated, especially in the zone located between the Mexico-Toluca highway and the Alzate Reservoir. Water samples and freshwater cladoceran were analysed for Cu and Zn by Energy Dispersive X-Ray Fluorescence (EDXRF) and for Hg and Cd by Neutron Activation Analysis (NAA). As a general feature, the heavy metal concentrations of the water were found to decrease in the sequence: Cu > Zn > Hg > Cd. Metallothioneins (MT) were determined by silver saturation method. Tissue concentrations of MT in Dhapnia pulex varied between 5.69 and 8.96 (mg MT/ g wet wt) in rain season and between 48.87 and 74.00 (mg MT/ g wet wt) in dry season. Metallothioneins levels in Dhapnia pulex were significantly correlated (P < 0.01) with tissue Hg concentrations. In contrast, correlations between MT and tissue levels of Cu and Zn were weak. These observations suggest that Hg2+ activity is the key environmental factor to which metallothionein levels in Daphnia pulex are responding in the studied reservoir

Experimental study of the ultraviolet global radiation in San Jose, Costa Rica

International Nuclear Information System (INIS)

Wright, J.

1996-01-01

The ultraviolet global radiation and the global solar radiation at San Jose, Costa Rica (latitude: 9 0 56', longitude: 84 0 54', altitude: 1.172 m.) during the period October 1993 to January 1995 were analyzed with respect to their seasonal variations and their independence. The dependence between the ultraviolet radiation and the clearness index of the skies was also investigated. A poor correlation was found between the quotient of the ultraviolet radiation (Hv/Hg) and between the global solar radiation and the extraterrestrial solar radiation (Hg/Ho). The correlation coefficient found between Hv/Hg and Hg/Ho was not greater than 0.25 for four categories of clearness index, i.e., covered skies, clear skies, and two intermediate conditions. This demonstrates that the ultraviolet radiation is not only associated with other atmospheric transmission conditions. A regression analysis between the hourly values of the ultraviolet and global radiation yielded a linear relationship with a determination coefficient greater than 98%. Thus a simple linear regression is reliable for the estimation of the ultraviolet in San Jose from global solar radiation data. (author) [es

Leaf anatomy of Gaillardia cabrerae (Asteraceae: Basic plan and comparative study of two contrasting habitat populations Anatomía foliar de Gaillardia cabrerae (Asteraceae: plan básico y estudio comparativo de dos poblaciones de hábitats contrastantes

Directory of Open Access Journals (Sweden)

Laura Beinticinco

2011-06-01

Full Text Available This study evaluates the leaf anatomy pattern of Gaillardia cabrerae Covas, an exclusive endemic camephyte from the Lihué Calel National Park, province of La Pampa, Argentina. Leaf cross sections and peelings of plants growing in two different populations were used to determine the basic leaf anatomy patterns and to estimate the influence of growth conditions in the two microhabitats. The analysis showed differences in epidermal cells area, stomata density and main diameter, lacunar parenchyma cells and central vascular bundle diameter and palisade parenchyma cells dimensions. These aspects might be reflecting environmental conditions of the populations, especially sun exposure and soil moisture. The results provide information on the relationship of the species to its environment, which could be used in the establishment of conservation policies.Gaillardia cabrerae Covas es un caméfito endémico del Parque Nacional Lihué Calel. En este estudio se evalúan características anatómicas de individuos de dos poblaciones provenientes de diferentes microhábitats de las Sierras de Lihué Calel. La información proporcionada es importante para entender las condiciones de vida de la especie y provee información inherente para establecer planes de conservación. Mediante cortes transversales y la técnica de peeling fue posible determinar el plan básico de la anatomía foliar y también verificar diferencias significativas en el área de células epidérmicas, densidad estomática, diámetro mayor de los estomas, diámetro de células del parénquima lagunar y del haz vascular central y en las dimensiones de las células del parénquima en empalizada. Las condiciones ambientales a las cuales estas poblaciones están sometidas, especialmente la exposición solar y la humedad del suelo son posibles factores que podrían explicar las diferencias detectadas en la anatomía foliar para ambas poblaciones.

Dating considerations about Mahoma Guaycuru complex (San Jose province-Uruguay)

International Nuclear Information System (INIS)

Medina, E.

1988-01-01

The aim of this work has been achieved the training in the radimetric determination of age of rocks for those methods of K/Ar and Rb/Sr; the knowledge of this methodology and the different technical steps that it demands, will allow to achieve a better understanding of the geologic evolution have suffered the rocks of the Crystaline oldland in our country. The rocks selected for the work come from the San Jose province, of 4th Judicial section, well-known place as Sierra Mahoma and Guaycuru.

Investigation for calculation methods used in analyzing the physics characteristics of nuclear power reactor

International Nuclear Information System (INIS)

Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong

2014-01-01

The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)

Geological prospecting in La Primera area (Caballo Torcido) Fotoplano M-25 Mal Abrigo San Jose town

International Nuclear Information System (INIS)

Coronel, N.; Techera, J.; Pias, J.; Ramos, E.; De Souza, S.

1990-01-01

This work describes the prospection results in La Primera (Caballo Torcido) area in San Jose town. In this zone has been extracted metallic minerals, sediments and multielement soil samples to detect geochemical anomalies.

The Great War and Remembrance in Jose Leon Machado's "Memoria das Estrelas sem Brilho"

Science.gov (United States)

Azevedo, Milton M.

2011-01-01

This article analyzes Jose Leon Machado's novel, "Memoria das Estrelas sem Brilho," as a multilayered historical novel in which a war story provides a background for comments on aspects of early twentieth-century Portuguese society, such as male bonding, religion, sexual mores, and social stratification. (Contains 11 notes.)

75 FR 71106 - San Jose Water Company; Notice of Application Accepted for Filing and Soliciting Comments...

Science.gov (United States)

2010-11-22

... consumptive domestic use. The hydraulic capacity of each turbine will be 9.5 cubic feet per second and the...-kilowatt turbine/ generator units in parallel with a pressure reduction valve at the San Jose Water Company...

Determination of Kr-85 in environmental samples and gaseous effluents from nuclear industries using the standard method

International Nuclear Information System (INIS)

Heras Iniquez, M.C.; Perez Garcia, M.M.

1983-01-01

The determination of the Kr-85 activity in environmental samples and gaseous effluents from Spanish Nuclear Power Stations is described. The method employed has been published elsewhere. The determinations has been carried out in environmental samples token at JEN Laboratories (Madrid) and the Nuclear Power Stations, Jose Cabrera (Zorita), Garona and Vandellos. Also samples of gaseous effluents of the three plants has been analyzed. Values of the Kr-85 environmental background activity in the Almaraz Nuclear Power Stations, has been determined, before the beginning of its nuclear activity. In this paper the sampling equipment used is described and the values found of Kr-85 activity in all the samples in given. (Author) 29 refs

Determination of Kr-85 in environmental samples and gaseous effluents from nuclear industries using the standard method; Aplicacion del metodo de concentracion separacion y medida radiactiva por centelleo liquido de Kr-85 en muestras ambientales y en efluentes gaseosos de la industria nuclear

Energy Technology Data Exchange (ETDEWEB)

Heras, M c; Perez, M M

1983-07-01

The determination of the Kr-85 activity in environmental samples and gaseous effluents from Spanish Nuclear Power Stations is described. The method employed has been published elsewhere. The determinations has been carried out in environmental samples token at JEN Laboratories (Madrid) and the Nuclear Power Stations, Jose Cabrera (Zorita), Garona and Vandellos. Also samples of gaseous effluents of the three plants has been analyzed. Values of the Kr-85 environmental background activity in the Almaraz Nuclear Power Stations, has been determined, before the beginning of its nuclear activity. In this paper the sampling equipment used is described and the values found of Kr-85 activity in all the samples in given. (Author) 29 refs.

The environmental thinking of Jose Bonifacio de Andrada e Silva

Directory of Open Access Journals (Sweden)

Maurecir Guimarães de Moraes

2015-02-01

Full Text Available The study aims at emphasizing the strong presence of the renowned Patriarch of Independence Jose Bonifacio de Andrada e Silva in discussions on environmental degradation that occurred in Brazil in the nineteenth century. The author registered his concern about deforestation in the Brazilian rainforests for farming and charcoal production, and warned about the depletion of water resources, as well as the disappearance of important trees – important raw-material for the shipbuilding industry. He proposed replanting seedlings, the rational use of natural resources, and recommended the adoption of agricultural practices only in valleys. These proposals show an aspect of Bonifacio that is unknown to Brazilian historiography.

Measurements of Po214 at Sao Jose dos Campos and Cachoeira Paulista: correlation with meteorological data

International Nuclear Information System (INIS)

Marinho, E.V.A.; Nordemann, D.J.R.

1985-01-01

PO 214 , a short life daughter product of RN 222 , was measured by alpha-ray spectrometry of a atmospheric particles sampled on Millipore filters (0.8 μm pore) at Sao Jose dos Campos (23 0 12'S, 45 0 51'W) and Cachoeira Paulista (22 0 41'S, 45 0 00'W). The infleunce of rainfall on the atmospheric Po 214 was studied at Sao Jose dos Campos. The mean activity was 98 pCi/m 3 for dry days (without rain during the preceding two days). On the contrary a low mean activity of 35 pCi/m 3 was observed for high pluviometry days from 08/30 to 10/26 for which total rainfall was 365.5 mm. At Cachoeria Paulista the Po 214 activity was correlated with the stability of the air (measured by the Richardson number), showing an accumaltion of natural radioactive aerosols during periods of higher stability of the lower atmosphere. (author) [pt

Memorias de alemanes en España durante la Guerra de la Independencia : la estancia de Philipp Schwein en la isla Cabrera

Directory of Open Access Journals (Sweden)

Hiltrud Friederrich-Stegmann

2003-01-01

Full Text Available El presente artículo tiene tres objetivos principales. Primero, da una información general de las memorias de alemanes participantes en la Guerra de la Independencia, junto con una relación bibliográfica, añadida como apéndice. En segundo lugar, da a conocer el autor alemán Johann Christian Mámpel que en seis libros, publicados anónimamente, cuenta las guerras napoleónicas desde el punto de vista del soldado raso, del Feldjáger. A Goetfie le gustó tanto esta obra que escribió el prólogo. Finalmente, el estudio incluye dos traducciones, la de un episodio del tercer tomo, que Mámpel dedica a la estancia de un compañero de guerra como prisionero en la desierta isla de Cabrera, y también la de la introducción de Goethe al mismo tomo con sus reflexiones sobre este asunto.The present article has three main objects. Firstly, it gives a general view of the memoirs written by Germán participants in the Spanish War of Independence, including an appendix with a related bibliography. Secondly, it introduces the German author Johann Christian Mámpel, who wrote six books, anonimously published, about the Napoleonic Wars from the point of view of the prívate soldier, the Feldjáger. Goethe was so impressed by it, that he wrote the prologue. Finally, the study includes two translations: an episode of the third volume, which Mámpel dedicates to a fellow soldier, who was a prisoner of war on the desert island of Cabrera, and also Goethe's introduction to the same volume with his reflections about this event.

Petrographic characterization, geochemistry and age of the nepheline syenite of San Jose del Guaviare

International Nuclear Information System (INIS)

Arango Mejia, Maria Isabel; Zapata Garcia, Gilberto; Martens, Uwe

2012-01-01

The San Jose del Guaviare Nepheline Syenite includes two rock bodies that crop out 20 km SW of the town with the same name. The first of these bodies, which had already been reported by Vesga and Castillo (1972), is exposed in the La Pizarra area. The second body, which had not been previously reported, occurs in the Las Delicias area.Petrographic characterization shows that rocks correspond to nepheline syenite, nepheline monzosyenite, and feldspar + nepheline granofels. The main minerals in the rocks are microcline + nepheline + plagioclase + -- biotite + - arfvedsonite, and accessory phases include fluorite, sphene, apatite, garnet, zircon, epidote and calcite. Structures and textures are chiefly igneous in the la Pizarra body and mainly metamorphic (restitic) in the Las Delicias body. The geochemical character of the rocks is alkaline, and they correspond to within- the field of rocks formed in intraplate environment. The above features suggest that the San Jose del Guaviare syenite was formed by anatexis in a continental environment.U/Pb dating of zircon and 40Ar/39Ar (HS) dating of biotite indicates age 577.8+ -/0 6.3 - 9 Ma (Ediacaran) crystallization and 494+ - 5 Ma (late Cambrian) cooling.

Jose Maria Albareda (1902-1966) and the Formation of the Spanish Consejo Superior de Investigaciones Cientificas

Science.gov (United States)

Malet, Antoni

2009-01-01

Jose Maria Albareda (1902-1966) was an applied chemist and a prominent member of the Roman Catholic organization, Opus Dei, who played a crucial role in organizing the Consejo Superior de Investigaciones Cientficas (CSIC), the new scientific institution created by the Franco regime in 1939. The paper analyses first the formative years in…

Identifying Population Vulnerable to Extreme Heat Events in San Jose, California.

Science.gov (United States)

Rivera, A. L.

2016-12-01

The extreme heat days not only make cities less comfortable for living but also they are associated with increased morbidity and mortality. Mapping studies have demonstrated spatial variability in heat vulnerability. A study conducted between 2000 and 2011 in New York City shows that deaths during heat waves was more likely to occur in black individuals, at home in census tracts which received greater public assistance. This map project intends to portray areas in San Jose California that are vulnerable to extreme heat events. The variables considered to build a vulnerability index are: land surface temperature, vegetated areas (NDVI), and people exposed to these area (population density).

Micro paleontology of Camacho, Raigon and Libertad formations in San Jose department in Uruguay

International Nuclear Information System (INIS)

Da Silva, J.

1990-01-01

Camacho, Raigon and Libertad formations of Miocene to Pleistocene age, are rich in microfossils as showed by recent studies of samples coming from cliffs at San Jose department, at Rio de la Plata shore and from Borehole N 1394/4 (DINAMIGE) which was made near to Kiyu resort. The microfossils are integrated by foraminiferids, diatoms, ostracods, silicophitoliths and uniaxial sponge spicles. By means of paleontologic analysis we can reach to a much clear definition of the geological evolution of the area

Vegetable output and cost savings of community gardens in San Jose, California.

Science.gov (United States)

Algert, Susan J; Baameur, Aziz; Renvall, Marian J

2014-07-01

Urban dwellers across the United States increasingly access a variety of fresh vegetables through participation in neighborhood-level community gardens. Here we document vegetable output and cost savings of community gardens in the city of San Jose, CA, to better understand the capacity of community gardens to affect food affordability in an urban setting. A convenience sample of 83 community gardeners in San Jose completed a background survey during spring and summer 2012. On average, gardeners were aged 57 years and had a monthly income of $4,900; 25% had completed college. A representative subset of 10 gardeners was recruited to weigh vegetable output of their plots using portable electronic scales at three separate garden sites. Accuracy of each portable scale was verified by comparing the weight of a sample vegetable to weights obtained using a lab scale precise to 0.2 oz. Garden yields and cost savings were tabulated overall for each plot. Results indicate that community garden practices are more similar to biointensive high-production farming, producing 0.75 lb vegetables/sq ft, rather than conventional agricultural practices, producing 0.60 lb/sq ft. Gardens produced on average 2.55 lb/plant and saved $435 per plot for the season. Results indicate that cost savings are greatest if vertical high value crops such as tomatoes and peppers are grown in community gardens, although yields depend on growing conditions, gardener's skill, availability of water, and other factors. Future research is needed to document cost savings and yields for specific crops grown in community gardens. Copyright © 2014 Academy of Nutrition and Dietetics. Published by Elsevier Inc. All rights reserved.

Engineering activities for the preparation of systems and facilities in the dismantling of Jose Cabrera NPP

International Nuclear Information System (INIS)

Gomez Rodriguez, C. A.; Martin Palomo, N.

2012-01-01

This paper presents the previous work of analysis of management systems and facilities, modifications to systems plans and the final implementation carried out on the site. The final result of the development of these plans, obtained after two years of intense work and in particular the result of the evolution of the turbine (now EAD) building, converted into the central infrastructure for the dismantling of the rest of the installation plans will be presented.

78 FR 37791 - In the Matter of: Jose Guadalupe Reyes-Martinez, Inmate Number #85993-279, CI Adams County...

Science.gov (United States)

2013-06-24

... DEPARTMENT OF COMMERCE Bureau of Industry and Security In the Matter of: Jose Guadalupe Reyes... the Internal Security Act of 1950 (50 U.S.C. 783(b)), or section 38 of the Arms Export Control Act (22... 750.8 of the Regulations states that the Bureau of Industry and Security's Office of Exporter Services...

Practical training in the operation of nuclear power plants with Interactive Graphic Simulator of Zorita; Formacion practica en la operacion de centrales nucleares con el Simulador Grafico Interactivo de Zorita

Energy Technology Data Exchange (ETDEWEB)

Cuervo, D.; Garcia-Herranz, N.; Garcia, S.; Davila, R.; Ahnert, C.; Aragones, J. M.; Cabellos, O.; Gallego, E.; Lorente, A.; Minguez, E.; Rebollo, L.; Blanco, J.

2010-07-01

In April 2008 a collaboration agreement was signed between Gas Natural Union Fenosa and the Universad Politecnica de Madrid for the creation of the Aula Jose Cabrera dedicated to train professionals in the field of nuclear technology. The Classroom located in the Department of Nuclear Engineering, has been equipped with the Interactive Graphic Simulator of Zorita (SGIZ). The use of the simulator intended to improve the quality of teaching in the area of Nuclear Engineering. It integrates in the teachings of Industrial Engineering degree and the Master of Nuclear Science and Technology. Different manuals are under preparation to make it a suitable tool for teaching purpose. These manuals will guide the student so that learning takes place both through the guidance of the teacher as independently. (Author) 3 refs.

Practical training in the operation of nuclear power plants with Interactive Graphic Simulator of Zorita

International Nuclear Information System (INIS)

Cuervo, D.; Garcia-Herranz, N.; Garcia, S.; Davila, R.; Ahnert, C.; Aragones, J. M.; Cabellos, O.; Gallego, E.; Lorente, A.; Minguez, E.; Rebollo, L.; Blanco, J.

2010-01-01

In April 2008 a collaboration agreement was signed between Gas Natural Union Fenosa and the Universad Politecnica de Madrid for the creation of the Aula Jose Cabrera dedicated to train professionals in the field of nuclear technology. The Classroom located in the Department of Nuclear Engineering, has been equipped with the Interactive Graphic Simulator of Zorita (SGIZ). The use of the simulator intended to improve the quality of teaching in the area of Nuclear Engineering. It integrates in the teachings of Industrial Engineering degree and the Master of Nuclear Science and Technology. Different manuals are under preparation to make it a suitable tool for teaching purpose. These manuals will guide the student so that learning takes place both through the guidance of the teacher as independently. (Author) 3 refs.

78 FR 37793 - In the Matter of: Jose Arturo Ramon-Herrada, Inmate Number #90903-279, CI Willacy County...

Science.gov (United States)

2013-06-24

... DEPARTMENT OF COMMERCE Bureau of Industry and Security In the Matter of: Jose Arturo Ramon-Herrada... Secretary of Commerce for Industry and Security. The appeal must be filed within 45 days from the date of....S.C. 793, 794 or 798; section 4(b) of the Internal Security Act of 1950 (50 U.S.C. 783(b)), or...

Lithofacies characterization related to the Raigon Formation located in the southern area of the San Jose Department

International Nuclear Information System (INIS)

Spoturno, J.; Morales, E.; Cazaux, S.; Aubet, N.; Loureiro, J.

2004-01-01

In this work the different lithofacies characterization related to the Raigon Formation, located in the southern area of the San Jose Department, is exposed supported by surface and subsurface data. Six stratigraphical sections were constructed considering lithological borehole descriptions to the aim of making a contribution on the spatial distribution, thickness, disposition, lithofaciological variations of this Formation and its stratigraphical relationships with other units [es

Jose Vasconcelos--A Man for All the Americas. The Tinker Pamphlet Series for the Teaching of Mexican American Heritage.

Science.gov (United States)

Miller, Hubert J.

Due to his dissatisfaction with the Diaz government, Jose Vasconcelos joined the revolutionary leader Francisco Madero, who along with other rebel leaders, brought an end to the Diaz regime in 1911. Vasconcelos shared the successes and misfortunes that followed Diaz's overthrow. When the Madero government came to an abrupt end in 1913, Vasconcelos…

Stock of immobilized metals in 'Jose Antonio Alzate' dam mud sediment samples in the State of Mexico, Mexico

International Nuclear Information System (INIS)

Rodriguez, S.A.; Avila-Perez, P.

1997-01-01

'Jose Antonio Alzate' dam mud sediment samples were collected from six sampling sites. The samples were analyzed by combining multielemental studies with structural details, including heating to very high temperatures. Characterization of mud sediment samples have been shown to have low concentrations of metals (Cr, Mn, Ni, Cu, Zn, Pb). No systematic trend was observed as a function of the sampling points. (author)

The short half-life descendents from radon measured in Sao Jose dos Campos and Cachoeira Paulista, Brazil, and its correlation with meteorological data

International Nuclear Information System (INIS)

Andrade Marinho, E.V. de.

1985-09-01

The correlations between the activity of the short half-life radon decay products in the low atmosphere and the meterological parameters, measuring the atmospheric polonium 214 are studied. The aerosols were collected on membrane filters and analysed by alpha spectrometry. Measurements have been made in Sao Jose dos Campos and Cachoeira Paulista, in Brazil. At S. Jose dos Campos the influence of the pluviometry on the concentration of atmospheric Po 214 has been analysed, showing that high activity correspond to low pluviometry and inversely, low activities correspond to high pluviometry. At Cachoeira Paulista, the variation in the atmospheric Po 214 activity in corresation with the air stability, measured, showing the accumulation of radioactive aerosols during the greater stability in the the lower atmospheric layers, was studied. (author) [pt

Jose de la Vega: a naval doctor who authored the first Spanish book on radiology

International Nuclear Information System (INIS)

Villanueva, A. J.; Rozua, L.; Garzon, M.; Ramirez, J. M.; Escatllar, T.

1999-01-01

Jose de la Vega Elorduy was a doctor in the Spanish Naval Medical Corps who wrote the book entitled Los Rayos X. Memoria sobre Radioscopia y Radiografia in 1899. We have learned that this book, published in Cartagena in 1900, was the first text about Radiology to be published in Spain, rather than that written by Major J. Mitjavila of the Army Medical Corps, as was previously thought. We review the book, stressing its pioneering outlook which provides brief view of the advent of Radiology in the Spanish Navy. (Author) 11 refs

Developmental pathway from leaves to galls induced by a sap-feeding insect on Schinus polygamus (Cav.) Cabrera (Anacardiaceae).

Science.gov (United States)

Dias, Graciela G; Ferreira, Bruno G; Moreira, Gilson R P; Isaias, Rosy M S

2013-03-01

Galling sap-feeding insects are presumed to cause only minor changes in host plant tissues, because they usually do not require development of nutritive tissues for their own use. This premise was examined through comparison of the histometry, cytometry and anatomical development of non-galled leaves and galls of Calophya duvauae (Scott) (Hemiptera: Calophyidae) on Schinus polygamus (Cav.) Cabrera (Anacardiaceae). Cell fates changed from non-galled leaves to galls during the course of tissue differentiation. C. duvauae caused changes in dermal, ground, and vascular systems of the leaves of S. polygamus. Its feeding activity induced the homogenization of the parenchyma, and the neoformation of vascular bundles and trichomes. The histometric and cytometric data revealed compensatory effects of hyperplasia and cell hypertrophy in the epidermis, with hyperplasia predominating in the adaxial epidermis. There was a balance between these processes in the other tissues. Thus, we found major differences between the developmental pathways of non-galled leaves and galls. These changes were associated with phenotypic alterations related to shelter and appropriate microenvironmental conditions for the gall inducer. The nondifferentiation of a typical nutritive tissue in this case was compared to other non-phylogenetically related arthropod gall systems, and is suggested to result from convergence associated with the piercing feeding apparatus of the corresponding gall-inducer.

Consideraciones taxonómicas en especies de Senecio de Argentina Taxonomic considerations on Argentine species of Senecio

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Roberto D. Tortosa

2010-12-01

Full Text Available Se incluyen algunas especies de Senecio de Argentina, reconocidas hasta el presente como especies válidas, en la sinonimia de otras: S. francisci Phil. en S. polygaloides Phil.; S. gymnocaulos Phil. en S. kingii Hook. f.; S. molinae Phil. en S. glaber Less.; S. mustersii Speg. y su var. dentatus Cabrera en S. subumbellatus Phil. Senecio microcephalus Phil. y su var. angustifolius Cabrera son consideradas una variedad de S. subumbellatus y S. schreiteri Cabrera una variedad de S. leucostachys Baker. No se acepta la validez de las variedades Senecio gilliesii Hook. et Arn. var. dasycarpus Cabrera, S. glaber Less. var. pratensis (Phil. Cabrera, S. gnidioides Phil. var. gilvus (Phil. Cabrera, S. kingii Hook. f. var. paradoxus (Albov ex Kurtz Cabrera, S. leucomallus A. Gray var. incisus A. Gray, S. linariifolius Poepp. ex DC. var. heliophytoides (Phil. Reiche y var. subtomentosus Cabrera y de la forma S. perezii Cabrera fo. integerrimus Cabrera.

Skills management medical labor in the plan of dismantling and decommissioning of the nuclear power plant Jose Cabrera

International Nuclear Information System (INIS)

Garcia Martinez, M.

2012-01-01

The model adopted for managing the skills of workers in the various contracts present in the dismantling in order to minimize, on the one hand, workplace accidents attributable to previous health status of the worker and the other the effects on the health could have a possible exposure to ionizing radiation. This model is based on the close coordination between the departments directly involved.

TECHNOLOGICAL CLUSTERING IN SÃO JOSE DOS CAMPOS: TRAJECTORIES AND IMPLICATIONS

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Isabel Cristina dos Santos

2013-08-01

Full Text Available This article analyses the relationship among entrepreneurship, innovation and growth path regarding the technological cluster in São Jose dos Campos, state of São Paulo. This is an exploratory and descriptive study, supported by documental research and questionnaires applied in a non-probabilistic sample. The results show a relevant social reorganization since CTA/ITA were implemented, in the early Fifties’, which coincides with the arrival of international companies, attracted by the Geographic aspects, high qualified manpower availability and tax benefits. Since then, the municipality has experienced an intense migratory flow of professionals, professors, civil and military students. CTA and ITA generate a considerable level of technological knowledge spillover. Due to it, a number of new companies have started-up operations locally in fields such as Aeronautics, Airspace, Automobile and Petrochemical, which promoted a remarkable social and economic development. Nevertheless, the small and medium technology-based enterprises seem to show high dependence of the core companies in the region that leads to initial stages of internationalization.

The Department of Energy/American Chemical Society Summer School in Nuclear and Radiochemistry at San Jose State University

International Nuclear Information System (INIS)

Kinard, W.F.; Silber, H.B.

2005-01-01

A Summer School in Nuclear Chemistry sponsored by the U. S. Department of Energy and the American Chemical Society has been held at San Jose State University for the past 20 years. The intent of the program is to introduce outstanding college students to the field of nuclear and radiochemistry with the goal that some of these students will consider careers on nuclear science. The program features radiochemistry experiments along with radiation safety training, guest lectures by well known nuclear scientists and field trips to nuclear chemistry facilities in the San Francisco area. (author)

El Filibusterismo and Jose Rizal as “Science Fictionist”

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Miguel Paolo P. Reyes

2013-12-01

Full Text Available Jose Rizal’s El Filibusterismo famously depicts the diff iculties faced by adherents of scientif ic thought in the Philippines during the late nineteenth century. It also contains descriptions of various implements that were uncommon for the time, from a “time bomb” within a lamp to a “specter summoner” that apparently projects hologram-like images—the products of reasonable extrapolations from the known science of the age. These features are common among the forerunners of the modern Anglo-American genre of science fiction, from Mary Shelley’s Frankenstein to a number of works by Edgar Allan Poe. It is far from the objective of this paper to push for Rizal’s canonization as the Father of Filipino Science Fiction, however. Instead, this paper seeks to surface the particular ways Rizal chose to depict the conflicts between knowledge borne of materialist investigation and knowledge from the earthly emissaries of the divine. It argues that in El Filibusterismo, these conflicts are somewhat helpful but at the same time dismissible as irrelevant in the context of a campaign against injustice—helpful in intimidating enemy oppressors, irrelevant among advocates of the downtrodden. In this way, Rizal’s novel implicitly contemplates a way to craft “science fiction” that strays from the imperialistic “science conquers” formula predominant in the West.

Learning from eponyms: Jose Verocay and Verocay bodies, Antoni A and B areas, Nils Antoni and Schwannomas

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Rajiv Joshi

2012-01-01

Full Text Available Schwannomas are benign peripheral nerve sheath neoplasms composed almost entirely of Schwann cells and are diagnosed histopathologically by the presence of singular architectural patterns called Antoni A and Antoni B areas. These were described first in 1920 by the Swedish neurologist Nils Antoni. The Antoni A tissue is highly cellular and made up of palisades of Schwann cell nuclei, a pattern first described in 1910 by the Uruguayan neuro-pathologist Jose Verocay and are known as Verocay bodies. This article describes the structure and appearance of Verocay bodies and Antoni A and B areas with a brief biographical introduction of the men who described these patterns.

Atividade antioxidante e toxicidade preliminar do extrato e frações obtidas das folhas e cascas do caule de Dasyphyllum tomentosum (Spreng. Cabrera

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C.S. Paula

2014-06-01

Full Text Available Dasyphyllum tomentosum (Spreng. Cabrera, açucará ou espinho-de-agulha, pertence à família Asteraceae, a qual compreende muitas espécies com propriedades terapêuticas. O objetivo deste trabalho foi avaliar extratos e frações de folhas e de cascas do caule de D. tomentosum, com relação as atividade antioxidante, citotóxica e hemolítica em testes in vitro. Todas as amostras apresentaram atividade antioxidante pelo método de inibição de DPPH, com destaque para a fração acetato de etila obtida das folhas cuja atividade foi comparável à dos padrões ácido ascórbico e rutina. Com relação à redução do complexo fosfomolibdênio, observou-se que esta mesma fração foi semelhante somente a rutina enquanto a fração obtida das cascas do caule apresentou resultado superior. Não foi observada atividade citotóxica e hemolítica frente aos modelos utilizados com os extratos e frações. Os resultados obtidos demonstram o potencial antioxidante da espécie sem apresentar toxicidade.

Estudio preliminar de las mineralizaciones de la mina de oro romana de Llamas de Cabrera (Leon, NO de España

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Cifuentes, J.

2005-12-01

Full Text Available The mine of Llamas de Cabrera (Leon, Spain is a gold deposit operated during the Roman Empire, in which primary and secondary gold ores coexist. The primary gold ores were hosted mainly in the upper layers of the Armorican Quartzite. They are lodes, being very abundant those hosted in N180°E/70-85°E tensional fractures and rare the ones associated to N70°E and N125°E fractures. They are formed by quartz and arsenopyrite very dominant over pyrite, galena, gold, chalcopyrite, bismuthinite, tetrahedrite and covellite. The accompanying hydrothermal alteration is silicification, sericitization and chloritization. The determined supergene minerals are scorodite, goethite and anglesite. The operated gold is free gold located generally in microfractures in arsenopyrite or quartz. Its content in silver varies from 1,99 to 7,62%, whereas the copper content is very low, remaining below the limit of instrumental detection. This study shows the existence of two phases of mineralization, deposited during two hydrothermal events of hercinian age, in addition to the supergenic phase. The secondary gold ores, operated in alluvial and colluvial sediments, were deployed by the action of the exogenous cycle on the primary gold deposits. We consider that the total volume of material moved at roman time ascended to 2 Mm3.La mina de Llamas de Cabrera (León, España es un yacimiento de oro explotado en época romana, en el que coexisten mineralizaciones de oro primario y de oro secundario. Las mineralizaciones de oro primario arman sobre todo en los tramos superiores de la Cuarcita Armoricana. Se trata de mineralizaciones filonianas, siendo muy abundantes las que encajan en fracturas tensionales de dirección N180°E/70-85°E y ocasionales las asociadas a fracturas de direcciones N70°E y N125°E. Están formadas por cuarzo y arsenopirita muy dominante sobre pirita, galena, oro, calcopirita, bismutina, tetraedrita y covellina. Las alteraciones

Anatomical and phenological implications of the relationship between Schinus polygama (Cav.) (Cabrera) and the galling insect Calophya rubra (Blanchard).

Science.gov (United States)

Guedes, L M; Aguilera, N; Ferreira, B G; Becerra, J; Hernández, V; Isaias, R M S

2018-05-01

The success of galling insects could be determined by synchronisation with host plant phenology and climate conditions, ensuring suitable oviposition sites for gall induction and food resources for their survival. The anatomical, histochemical and phenological synchronisation strategies between Calophya rubra (Blanchard) (Hemiptera: Psylloidea) and its host, the evergreen plant Schinus polygama (Cav.) (Cabrera) (Anacardiaceae), in the Mediterranean climate of southern Chile was evaluated and compared to that of the congeneric C. cf. duvauae (Scott) from Brazil and closely related host plant S. engleri in a subtropical climate. Anatomical, histometric, histochemical and vegetative phenology studies of the stem and galls were conducted from June 2015 to December 2016. Based on the anatomical, histometric and histochemical analysis, the conical stem gall traits imply gains over the non-galled stem toward the galling insect survival, but the maintenance of phellem, secretory ducts and pith indicate conservative developmental traits that cannot be manipulated by C. rubra. Our results indicate that the conditions of the Mediterranean climate zone limit C. rubra immature activity during unfavourable periods, probably determining a diapause period and a univoltine life cycle, which are peculiarities of the S. polygama- C. rubra system. The synchronisation between development and seasonality confers peculiarities to the S. polygama- C. rubra system in the Mediterranean climate zone. © 2018 German Society for Plant Sciences and The Royal Botanical Society of the Netherlands.

Report on the mineral exploration in the San Jose and Arroyo Grande area Oriental Republic of Uruguay

International Nuclear Information System (INIS)

2001-01-01

This survey is based on the scope of work signed between the Japanese government and the Oriental Republic of Uruguay on 24 November 2000. The purpose of this is to clarify both the geologic appearance and the occurrence of the ore fields of mineral deposits in the San Jose and Arroyo Grande area in this country, with the aim of discovering new ore deposits. In addition, another purpose is to transfer the technology to the involved organizations of the object country. The survey is conceived as a three year project initiated in 2000, and this fiscal year falls on the first phase.This survey consist of this existing data analysis, the geological interpretation of satellite image data, the geological survey and geochemical prospecting

Report on the mineral exploration in the San Jose and Arroyo Grande area Oriental Republic of Uruguay

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This survey is based on the scope of work signed between the Japanese government and the Oriental Republic of Uruguay on 24 November 2000. The purpose of this is to clarify both the geologic appearance and the occurrence of the ore fields of mineral deposits in the San Jose and Arroyo Grande area in this country, with the aim of discovering new ore deposits. In addition, another purpose is to transfer the technology to the involved organizations of the object country. The survey is conceived as a three year project initiated in 2000, and this fiscal year falls on the first phase.This survey consist of this existing data analysis, the geological interpretation of satellite image data, the geological survey and geochemical prospecting

Solar-energy-system performance evaluation. San Anselmo School, San Jose, California, April 1981-March 1982

Energy Technology Data Exchange (ETDEWEB)

Pakkala, P.A.

1982-01-01

The San Anselmo School is a one-story brick elementary school building in San Jose, California. The active solar energy system is designed to supply 70% of the space heating and 72% of the cooling load. It is equipped with 3740 square feet of evacuated tube collectors, a 2175-gallon tank for heat storage, a solar-supplied absorption chiller, and four auxiliary gas-fired absorption chillers/heaters. The measured solar fraction of 19% is far below the expected values and is attributed to severe system control and HVAC problems. Other performance data given for the year include the solar savings ratio, conventional fuel savings, system performance factor, and solar system coefficient of performance. Also tabulated are monthly performance data for the overall solar energy system, collector subsystem, space heating and cooling subsystems. Typical hourly operation data for a day are tabulated, including hourly isolation, collector array temperatures (inlet and outlet), and storage fluid temperatures. The solar energy use and percentage of losses are also graphed. (LEW)

Solar-energy system performance evaluation. San Anselmo School, San Jose, California, July 1980-March 1981

Energy Technology Data Exchange (ETDEWEB)

Pakkala, P.A.

1981-01-01

The San Anselmo School is a one-story, brick elementary school building located in San Jose, California. The active solar energy system is designed to supply 70% of the heating load and 72% of the cooling load. It is equipped with 3.740 square feet of evacuated tube collectors, 2175-gallon tank for storage, four auxiliary gas-fired absorption chiller/heaters, and a solar-supplied absorption chiller. The measured heating and cooling solar fractions were 9% and 19%, respectively, for an overall solar fraction of 16%, the lowered performance being attributed to severe system control problems. Performance data include the solar savings ratio, conventional fuel savings, system performance factor, and solar system coefficient of performance. Performance data are presented for the overall system and for each subsystem. System operation and solar energy utilization data are included. Also included are a description of the system, performance evaluation techniques, sensor technology, and typical performance data for a month. Weather data are also tabulated. (LEW)

The Sao Jose do Rio Pardo mangeritic-granitic suite, south eastern Brazil

International Nuclear Information System (INIS)

Campos Neto, M.C.; Figueiredo, M.C.H.; Janasi, V.A.; Basei, M.A.S.; Fryer, B.J.

1988-01-01

In the Sao Jose do Rio Pardo region, Sao Paulo and Minas Gerais States, occur some intrusive, folded tabular bodies of mangerites associated with hornblende granitoids. The country rocks correspond to a complex association of gneisses and migmatites, locally with granulite facies assemblages. Both the magnerites and hornblende granitoids present a tectonic foliation with mineral flattening and stretching. Petrographically the mangeritic rocks are mainly dark green quartz mangerites with mesoperthite, plagioclase, quartz, hypersthene, clinopyroxene and variable amounts of hornblende, with zircon as conspicuous acessory. The pink hornblende granitoids are mainly granite s.s. exhibiting higher quartz and amphibole contents and lacking pyroxenes. Hololeucocratic alkali feldspar granites are locally associated to the hornblende granites. The textures of the mangerites and granites almost always show an important metamorphic overprinting, with relictic mesoperthite and pyroxene crystal into a granoblastic matrix. The magneritic-granitic suite is characterized by relatively high Fe/(Fe + Mg), K and HFS elements and low Ca contents, being comparable to typical anorogenic magneritic-granitic suites from Svcandinavia and North America. The Rb/Sr data indicate a Late Proterozoic metamorphic isotopic rehomogenization (930 Ma, Ro = 0.706). Geological evidence suggest that the intrusive age could be Middle Proterozoic, wich is reinforced by another Rb-Sr value of about 1300 Ma. (author) [pt

Groundwater arsenic content in Raigon Aquifer System (San Jose, Uruguay)

International Nuclear Information System (INIS)

Manay, N.; Piston, M.; Goso, C.; Fernnandez, T.; Rejas, M.; Garcia Valles, M.

2013-01-01

As a Medical Geology research issue, an environmental arsenic risk assessment study in the most important sedimentary aquifer in southern Uruguay is presented. The Raigon Aquifer System is the most exploited in Uruguay. It has a surface extent of about 1,800 square kilometres and 10,000 inhabitants in San Jose Department, where it was studied. Agriculture and cattle breeding are the main economic activities and this aquifer is the basic support. The groundwater sampling was done on 37 water samples of PRENADER (Natural Resources Management and Irrigation Development Program) wells. Outcropping sediments of Raigon Formation and the overlying Libertad Formation were also sampled in the Kiyu region. The analyses were performed by inductively coupled plasma-optical emission spectrometry (ICP-OES) and inductively coupled plasma-mass spectrometry (ICP-MS). The results showed 80% samples with arsenic levels exceeding the 10 μg/l of WHO as limit for waters, and 11% exceeds the 20 μg/l limit of uruguayan regulation. The median, maximum and minimum water arsenic concentrations determined have been 14.24, 24.19 and 1.44 μg/l, respectively. On the other hand, nine sediment samples of Raigon and Libertad Formations in Kiyu region were analysed and yielded median, maximum and minimum arsenic concentrations of 5.03, 9.82 and 1.18 ppm, respectively. This issue leads to the supposition that the population, as well as industrial and agricultural activities, are consuming water with arsenic concentrations over the national and international maximum recommended limit.

Evaluation of authenticity and green tea antimicrobial activity of different commercial brands that are sold in San Jose, Costa Rica

International Nuclear Information System (INIS)

Alvarez Campos, Jeyson

2014-01-01

Various brands of green tea are compared with the leaves of camellia sinensis to verify the antimicrobial effect, authenticity and quality. The main components with antibacterial effect commercially presented are compared to the components present in green tea leaf to determine the quality and authenticity, by thin layer chromatography. The brands marketed studied green tea in the metropolitan area of San Jose, Costa Rica has been really green tea good quality, this can be ensured by the findings in the analysis and comparison of the different profiles of flavonoids, polyphenols, alkaloids and amino acids. The antimicrobial activity has been without to confirmed in the bands of catechins analyzed [es

Decree 343/012. Is regulated the servitude established by Decree-Law 10,383, on several lines of electricity conduction -150 KV to be built in the departments of Florida, Cerro Largo, Maldonado, Flores, San Jose and Tacuarembo

International Nuclear Information System (INIS)

2012-01-01

This decree regulates the establishment of electricity conduction in Florida, Cerro Largo, Maldonado, Flores, San Jose and Tacuarembo towns. These electrical lines are necessary to provide the public service by UTE

Evolution of the east-central San Jose del Cabo basin, Baja California Sur, Mexico

Science.gov (United States)

McTeague, M. S.; Umhoefer, P. J.; Schwennicke, T.; Ingle, J. C.; Cortes Martinez, M.

2006-12-01

The San Jose del Cabo basin at the southern tip of the Baja California peninsula records the early tectonic evolution of the west side of the Gulf of California. This study focused on the east central margin of the basin. The basal La Calera Formation unconformably overlies Cretaceous granite and consists of conglomerate, pebbly sandstone and conglomerate, and sandstone deposited in alluvial fans and fan-deltas. Deposition of the La Calera Formation was from ca. 9-14 Ma. The lower member of the Trinidad Formation was deposited beginning ca. 9-13 Ma and consists of sandstone, mudstone, and shelly mudstone deposited in nearshore and estuarine environments. These age estimates are based on sedimentation rates and foraminifera and coccoliths from the NN 11A nannozone (7.4 8.6 Ma, GTS 2004). The middle member of the Trinidad Formation consists of deeper water mudstones deposited by turbidity currents and suspension settling in a shelf to slope and conglomerates deposited by submarine debris flows on the shelf. The basin began earlier than previously thought. The oldest marine rocks are ca.9-13 Ma, while sedimentation on the east side began at ca. 9-14 Ma, synchronous with estimates of initiation of offset on the San Jose del Cabo fault. The Zapote fault is a down-to-the-east normal and sinistral-oblique fault that exposes a wedge of granite and older strata in the footwall to the west. The fault was active during sedimentation in the late Miocene and possibly later. The fault divides the study area into an eastern hanging wall subbasin and western footwall subbasin. The eastern subbasin formed an embayment in the eastern margin of the Cabo basin. A regional flooding surface (ca. 8 Ma) can be correlated across the fault that marks a major marine incursion. Depositional systems evolved rapidly from coarse-grained terrestrial systems to fine-grained marine and estuarine systems. The Cabo basin provides an excellent analogue for comparison with offshore basins, which are

Measurement of dose to skin using TLD of several radiodiagnostic studies in San Jose, Costa Rica; Medicion de dosis a piel utilizando TLD de varios estudios radiodiagnosticos en San Jose, Costa Rica

Energy Technology Data Exchange (ETDEWEB)

Mora, P. [Laboratorio de Fisica Nuclear Aplicada, Escuela de Fisica, Universidad de Costa Rica, San Jose (Costa Rica)

1998-12-31

It is quantified the radiation doses on skin for several radiodiagnostic studies in patients of the Calderon Guardia Hospital in San Jose, Costa Rica at the period October 1997-September 1998 using thermoluminescent dosemeters TLD 100. The crystals receive the decoction standard procedures and they are arranged at the middle of the irradiation field. For a total of 973 radiodiagnostic studies it was found that the dose on skin in mGy are: 2.09 for thorax AP/AP, 5.33 for thorax LAT, 5.35 for skull AP/PA, 2.98 for skull LAT, 10.74 for abdomen, hips and pelvis, 6.20 for spines AP, 9.35 for spines LAT, 11.48 for lumbar columns AP, 29.99 for lumbar columns LAT and 6.87 for intravenous skin diagrams (first plate ap). It is produced thus the first reference bank for the national hospitals, which is compared with the orientation levels of doses for IAEA. Recommendations to diminish the collective doses through quality control programs are discussed, taking as goal to have got radiographs of excellent diagnostic quality, but with the less possible doses. (Author)

Community and home gardens increase vegetable intake and food security of residents in San Jose, California

Directory of Open Access Journals (Sweden)

Susan Algert

2016-04-01

Full Text Available As of 2013, 42 million American households were involved in growing their own food either at home or in a community garden plot. The purpose of this pilot study was to document the extent to which gardeners, particularly less affluent ones, increase their vegetable intake when eating from either home or community garden spaces. Eighty-five community gardeners and 50 home gardeners from San Jose, California, completed a survey providing information on demographic background, self-rated health, vegetable intake and the benefits of gardening. The gardeners surveyed were generally low income and came from a variety of ethnic and educational backgrounds. Participants in this study reported doubling their vegetable intake to a level that met the number of daily servings recommended by the U.S. Dietary Guidelines. Growing food in community and home gardens can contribute to food security by helping provide access to fresh vegetables and increasing consumption of vegetables by gardeners and their families.

Remediation of mercury-polluted soils using artificial wetlands.

Science.gov (United States)

García-Mercadoa, Héctor Daniel; Fernándezb, Georgina; Garzón-Zúñigac, Marco Antonio; Durán-Domínguez-de-Bazúaa, María Del Carmen

2017-01-02

Mexico's mercury mining industry is important for economic development, but has unfortunately contaminated soils due to open-air disposal. This case was seen at two sites in the municipality of Pinal de Amoles, State of Queretaro, Mexico. This paper presents an evaluation of mercury dynamics and biogeochemistry in two soils (mining waste soil) using ex-situ wetlands over 36 weeks. In soils sampled in two former mines of Pinal de Amoles, initial mercury concentrations were 424 ± 29 and 433 ± 12 mg kg -1 in La Lorena and San Jose, former mines, respectively. Typha latifolia and Phragmites australis were used and 20 reactors were constructed (with and without plants). The reactors were weekly amended with a nutrient solution (NPK), for each plant, at a pH of 5.0. For remediation using soils from San Jose 70-78% of mercury was removed in T. latifolia reactors and 76-82% in P. australis reactors, and for remediation of soils from La Lorena, mercury content was reduced by 55-71% using T. latifolia and 58-66% in P. australis reactors. Mercury emissions into the atmosphere were estimated to be 2-4 mg m -2 h -1 for both soils.

Gamma dosimetry of the uranium deposit in Sao Jose de Espinharas, Paraiba, Brazil

International Nuclear Information System (INIS)

Silva, Alberto Antonio da; Santos Junior, Jose Araujo dos; Cunha, Andre Felippe Vieira da; Amaral, Romilton dos Santos; Oliveira, Iane Andrade de; Bezerra, Jairo Dias; Silva, Flavio Ferreira da

2013-01-01

Radioecology studies contribute to the monitoring of both anthropic and natural radionuclides, as well as their correlation with the ecosystem and the population. Terrestrial sources, which mainly include the primary radionuclides of the 238 U, 232 Th, and 40 K series, contribute to a higher effective dose received by humanity, with 84% from terrestrial radionuclides and another 16% derived from cosmogenic nuclides. Areas with high levels of natural radiation, above the acceptable limits, are of great relevance for the development of dosimetry studies, whose core aims include the preservation of the environment and the control of the population's exposure to radiation. The town of Sao Jose de Espinharas, Paraiba, Brazil, has a uranium oxide mine with an average grade of 1,200 mg/kg. In the present study, non-destructive, in situ assays were performed along the entire radiometrically anomalous area. The results obtained in terms of effective doses varied from 3.79 to 93.80 mSv.y -1 , with an average of 19.47 mSv.y -1 thus leading to the conclusion that all of the monitored points presented environmental doses of above the reference value of 2.4 mSv.y -1 , suggesting that both qualitative and quantitative analyses concerning the environmental matrixes in question need to be performed. (author)

Estudo florístico de segmentos de Mata Atlântica em Parelheiros, São Paulo, SP, para soltura de bugio-ruivo Alouatta clamitans (Cabrera, 1940. Floristic survey of Atlantic Rain Forest in the District of Parelheiros, São Paulo, southeastern Brazil, for reintroduction of the Brown Howler Monkey, Alouatta clamitans (Cabrera, 1940

Directory of Open Access Journals (Sweden)

Ricardo José Francischetti GARCIA

2014-06-01

Full Text Available A fim de minimizar os efeitos da pressão da urbanização sobre bugios-ruivos, Alouatta clamitans (Cabrera, 1940, a Divisão de Fauna em parceria com o Herbário Municipal realiza, desde 1996, o Programa de Reintrodução de Bugios, com o objetivo de viabilizar o retorno à vida livre dos indivíduos aptos. Entre várias áreas prospectadas para possível soltura de bugios-ruivos, o Sítio Roda d’Água, distrito de Parelheiros, município de São Paulo, foi selecionado para realização do presente estudo. Todos os componentes da mata, em uma área de 89,9 hectares, foram amostrados para avaliação da disponibilidade de alimentos. Foram registradas 251 espécies vasculares, sendo 13 pteridófitas, uma gimnosperma e 237 angiospermas. O material testemunho está depositado no Herbário Municipal. Na área estudada foram registradas duas espécies ameaçadas e quatro quase ameaçadas. As espécies dos componentes arbóreo-arbustivo e epifítico, que constituem o hábitat preferencial dos bugios-ruivos, correspondem a 73,7% do total levantado. Dentro deste grupo de espécies, 17,8% já foram registradas como fontes alimentares para a espécie-alvo. Dentre as espécies já referidas na dieta de bugios-ruivos, registradas na área de estudo, todas apresentam distribuição geográfica ampla. Desta forma, a área estudada foi considerada apropriada para a soltura de bugios-ruivos, quanto à disponibilidade de recursos para alimentação. In order to minimize the urbanization pressure effects over Brown Howler Monkeys, Alouatta clamitans (Cabrera, 1940, the Fauna Division in partnership with the Municipal Herbarium carries out a Brown Howler Monkey Reintroduction Program since 1996, with the aim of enabling the return of fit individuals to the wild. Among several areas prospected for possible reintroduction of the Brown Howler Monkey, Roda d’Água Ranch, District of Parelheiros, City of São Paulo, was selected to carry out this study. All

Lead-lead isotope method. Application to migmatites and associaded rocks of the region of Sao Jose dos Campos, SP

International Nuclear Information System (INIS)

Babinski, M.

1988-01-01

Establishment of chemical and analytical procedures, preparation and purification of reagents and the mounting of a clean laboratory (class 100) for the Pb-Pb isotope dating of whole rock and mineral samples are discussed in this work. Two different methods of dissolution of rock samples and chemical separation of lead using anionic resin were tried in a conventional chemical laboratory using analytical grade reagents. The samples analysed were paleossome and neossome of migmatites from Sao Jose dos Campos, SP, and from the results obtained the analytical method to be employed was chosen. The results obtained by this method agree well with those of the Geochronology Laboratory at Oxford, for the same samples. Applying the established analytical procedure, migmatites and associated gneisses from the region of Sao Jose dos Campos were dated by Pb/Pb method. Some of the samples were also dated by Rb-Sr method. The lead-lead isochron age of 2355 +- 85 M.a. (μ 1 = 8.33) for the paleossome may be interpreted as the age of the metamorphism for these rocks, which were formed probably during the Archean. The age data for the phacoidal gneisses show a similar geological evolution, whereas the 2414 +- 800 M.A. Pb-Pb isochron age for the protomylonitic gneisses indicates a disturbed Pb-Pb isotope system. The neossome samples define a Pb-Pb isochron age of 1388 +- 290 M.A. (μ1 = 8.15), which may be interpreted as the age of migmatization. The calculated Pb-Pb ages for two alkali feldspars from the migmatites are devoid of any geological significance. The age values obtained by the Rb-Sr isotope system for the neossome and paleossome of the migmatites agree with the Pb/Pb data within experimental errors. However, the study shows that results of the Rb-Sr data alone cannot be interpreted unambiguously with out comparison with data from other isotope systems. (author) [pt

Regional-scale impact of storm surges on groundwaters of Texas, Florida and Puerto Rico after 2017 hurricanes Harvey, Irma, Jose, Maria

Science.gov (United States)

Sellier, W. H.; Dürr, H. H.

2017-12-01

Hurricanes and related storm surges have devastating effects on near-shore infrastructure and above-ground installations. They also heavily impact groundwater resources, with potentially millions of people dependant on these resources as a freshwater source. Destructions of casings and direct incursions of saline and/or polluted waters have been widely observed. It is uncertain how extensive the effects are on underground water systems, especially in limestone karst areas such as Florida and Puerto Rico. Here, we report regional-scale water level changes in groundwater systems of Texas, Florida and Puerto Rico for the 2017 Hurricanes Harvey, Irma, Jose and Maria. We collected regional scale data from the USGS Waterdata portal. Puerto Rico shows the strongest increase in groundwater levels in wells during Hurricane Maria, with less reaction for the preceding storms Irma and Jose. Increases in water levels range from 0.5 to 11m, with maximum storm surges in Puerto Rico around 3m. These wells are located throughout Puerto Rico, on the coast and inland. In Florida, most wells that show a response during Hurricane Irma are located in the Miami region. Wells located on the west coast show smaller responses with the exception of one well located directly on Hurricane Irma's track. These wells show an increase of 0.2 to 1.7m. In Texas, wells located in proximity to Hurricane Harvey's track show an increase in water level. The effect of groundwater level increases is not limited to the Texas coast, but inland as well. An increase between 0.03 and 2.9m is seen. Storm surges for both Florida and Texas have ranged from 1.8-3.7m maximum. We discuss the findings in the context of local and regional geology and hydrogeology (presence of connected aquifer systems, faulting, presence of carbonate/karst systems etc.).

Bioaccumulation of chemical elements by water hyacinth (Eichhornia crassipes) found in 'Jose Antonio Alzate' dam samples in the State of Mexico, Mexico

International Nuclear Information System (INIS)

Rodriguez, S.A.; Avila-Perez, P.; Barcelo-Quintal, I.D.

1998-01-01

A study was undertaken to determine experimentally the uptake of pollutants into of the different parts of the water hyacinth (Eichhornia crassipes) found in 'Jose Antonio Alzate' dam in the State of Mexico, Mexico. There is evidence for efficient and significant root accumulation of Ti, Mn, Fe, and Ba; but in the upper parts concentrations was consistently determined by the degree of watering. However, a significant input could be derived from a common generic source, such as the atmospheric deposition. The experimental study would, therefore, indicate that water hyacinth species can be highly effective in providing a control and treatment buffer for toxic discharges to the dam. (author)

Behavior of the ionosphere total electronic content in Sao Jose dos Campos during magnetic storms in 1980

International Nuclear Information System (INIS)

Paula, E.R. de; Abdu, M.A.; Kantor, I.J.

1983-07-01

Faraday rotation data from 1980, obtained with a polarimeter at Sao Jose dos Campos (23 0 S, 46 0 W), were analyzed during periods occurring magnetic storms. In order to select these periods, the magnetic index Dst was used. It was observed that during magnetic storms preceeded by a few calm days, an increase in the Total Electron Content (TEC) is observed during the storm main phase, relative to the mean of the magnetic calm days (positive phase). Afterwards, during the storms recovery phase, a decrease was registered relative to the average (negative phase). This TEC behaviour, observed at low latitudes storms, is typical of the behaviour over medium latitudes. But, when several storms occur with few intervening days between them, the positive phase seems to prevail. This indicates an inibition of the source of the negative phase. This work discusses the possible origins of the positive and negative phases. (Author) [pt

Application of GIS and Visualization Technology in the Regional-Scale Ground-Water Modeling of the Twentynine Palms and San Jose Areas, California

Science.gov (United States)

Li, Z.

2003-12-01

Application of GIS and visualization technology significantly contributes to the efficiency and success of developing ground-water models in the Twentynine Palms and San Jose areas, California. Visualizations from GIS and other tools can help to formulate the conceptual model by quickly revealing the basinwide geohydrologic characteristics and changes of a ground-water flow system, and by identifying the most influential components of system dynamics. In addition, 3-D visualizations and animations can help validate the conceptual formulation and the numerical calibration of the model by checking for model-input data errors, revealing cause and effect relationships, and identifying hidden design flaws in model layering and other critical flow components. Two case studies will be presented: The first is a desert basin (near the town of Twentynine Palms) characterized by a fault-controlled ground-water flow system. The second is a coastal basin (Santa Clara Valley including the city of San Jose) characterized by complex, temporally variable flow components ­¦ including artificial recharge through a large system of ponds and stream channels, dynamically changing inter-layer flow from hundreds of multi-aquifer wells, pumping-driven subsidence and recovery, and climatically variable natural recharge. For the Twentynine Palms area, more than 10,000 historical ground-water level and water-quality measurements were retrieved from the USGS databases. The combined use of GIS and visualization tools allowed these data to be swiftly organized and interpreted, and depicted by water-level and water-quality maps with a variety of themes for different uses. Overlaying and cross-correlating these maps with other hydrological, geological, geophysical, and geochemical data not only helped to quickly identify the major geohydrologic characteristics controlling the natural variation of hydraulic head in space, such as faults, basin-bottom altitude, and aquifer stratigraphies, but also

21st Century cartography in the service of Enresa; Cartografia del siglo XXI al servicio de Enresa

Energy Technology Data Exchange (ETDEWEB)

Cienfuegos Hevia, I.

2013-05-01

Geographic information systems (GIS) are used to store and process both physical and social-economic information of a territory. Topography, vegetation, hydrology, geology, weather conditions, population, protected spaces. An endless list of details that, organized and put on maps, are very useful in numerous fields, including the handling of radioactive waste. Enresa uses these tools in many of its activities: site selection, characterization studies, monitoring of radiological and environmental surveillance programs in facilities such as El Cabril (Cordoba), or in the preparation of documentation that helped to locate the Centralized Temporary Storage (CTS) facility for high level waste in Villar de Canas (Cuenca). These cartographic instruments have also shown their usefulness in the dismantling of facilities such as the old Uranium Factory in Andujar (Jaen) or the nuclear power stations of Vandellos I (Tarragona), currently in latency phase, and Jose Cabrera (Zorita, Guadalajara), which is currently being decommissioned. At present, GIS tools provide major advantages: they reduce the number of reports and conventional map collections, support faster and easier access to information in more manageable ways resulting in technical resource and time savings, quick updates of maps and records under constant evolution, and streamlined decision-taking processes. (Author)

Education and training of future nuclear engineers through the use of an interactive plant simulator

International Nuclear Information System (INIS)

Ahnert, C.; Cuervo, D.; Garcia-Herranz, N.; Aragones, J.M.; Cabellos, O.; Gallego, E.; Minguez, E.; Lorente, A.; Piedra, D.; Rebollo, L.; Blanco, J.

2010-01-01

The International Atomic Energy Agency (IAEA) sponsors the development of nuclear reactor simulators for education, or arranges the supply of such simulation programs. Aware of this, the Department of Nuclear Engineering of the Universidad Politecnica de Madrid was provided in 2008 with the Interactive Graphical Simulator of the Spanish nuclear power plant Jose Cabrera, whose operation ceased definitively in 2006. According with the IAEA-TECDOC- 1411, the simulator is a Graphical Simulator, used for training of main control room personnel, technical support engineers, and operations management. This paper presents all the work performed at the Department to turn the simulator into a teaching/learning tool, to be use in the nuclear engineering studies following guidance found in: Shtub, A. Parush, T.T. Hewett 'The use of simulation in learning and teaching' (Int. J. Eng. Educ., 25(2), 2009, pp. 206-208). The experience obtained so far with the use of the simulator has been very successful. The graduate students involved in the development of the projects, practices and documents related with the simulator show a great interest for the work that they are doing making that the laboratory where the simulator is installed to be busy place. Regarding the undergraduate students, the practices in the simulator encourage them to follow the Nuclear Energy studies in the Engineering Schools, what is very rewarding for the Department professors. The simulator has proved to be an optimal tool to transfer the knowledge of the physical phenomena that are involved in the nuclear power plants, from the nuclear reactor to the whole set of systems and equipments on a nuclear power plant. It is also a relevant tool for motivation of the students, and to complete the theoretical lessons. This use of the simulator in the learning-teaching process meats also the criteria recommended for the Bologna adapted studies, as it helps to increase the private hands-on work of the student, and

Getting your home ready - knee or hip surgery

Science.gov (United States)

Hip or knee surgery - getting your home ready; Osteoarthritis - knee ... Cabrera JA, Cabrera AL. Total hip replacement. In: Frontera WR, Silver JK, Rizzo TD, eds. Essentials of Physical Medicine and Rehabilitation . 3rd ed. Philadelphia, PA: Elsevier Saunders; ...

Measurement of dose to skin using TLD of several radiodiagnostic studies in San Jose, Costa Rica

International Nuclear Information System (INIS)

Mora, P.

1998-01-01

It is quantified the radiation doses on skin for several radiodiagnostic studies in patients of the Calderon Guardia Hospital in San Jose, Costa Rica at the period October 1997-September 1998 using thermoluminescent dosemeters TLD 100. The crystals receive the decoction standard procedures and they are arranged at the middle of the irradiation field. For a total of 973 radiodiagnostic studies it was found that the dose on skin in mGy are: 2.09 for thorax AP/AP, 5.33 for thorax LAT, 5.35 for skull AP/PA, 2.98 for skull LAT, 10.74 for abdomen, hips and pelvis, 6.20 for spines AP, 9.35 for spines LAT, 11.48 for lumbar columns AP, 29.99 for lumbar columns LAT and 6.87 for intravenous skin diagrams (first plate ap). It is produced thus the first reference bank for the national hospitals, which is compared with the orientation levels of doses for IAEA. Recommendations to diminish the collective doses through quality control programs are discussed, taking as goal to have got radiographs of excellent diagnostic quality, but with the less possible doses. (Author)

Petrography and geochronology of the Sao Jose dos Quatro Marcos granitoids, southwestern Mato Grosso, Brazil

International Nuclear Information System (INIS)

Carneiro, M.A.; Ulbrich, H.H.G.J.; Kawashita, K.

1989-01-01

An area of about 400Km 2 was mapped in the region of Sao Jose dos Quatro Marcos, southwestern State of Mato Grosso, western Brazil. Various granitoid rocks were recognized, intrusive into basement gneisses, and mapped as two distinct units, one of massive late granitoids (mostly 3a and 3b granites) and one of foliated granitoids, with oriented mafic minerals, consisting mainly of tonalites and, to a lesser degree, of granodiorites. The sequence of intrusion is from tonalites to granodiorites to late granites. Magmatic textures are usually preserved in the rocks, with the exception of some clearly recrystallized samples. Field aspects, as well as petrography and mineralogy, suggest a genetic relationship between the various granitoid types, showing overall features commonly cited as typical of 'I' lineages. A Rb/Sr whole-rock isochron obtained from 3b granites (six points) shows an age of 1,472 +- 19 Ma, and Sr 87 86 initial ratio of 0.7037 +- 0.0004; a nine point isochron (adding two tonalites and one granodiorite) yelds similar results (age of 1,505 +- 20 Ma, initial ratio of 0.7029+- 0.0003). The obtained ages confirm the existence of a thermal-magmatic event at about 1,500 Ma at the southern border of the Amazon Craton. (author) [pt

Composição e dinâmica de grupos de Alouatta guariba clamitans Cabrera (Primates, Atelidae em Floresta Ombrófila Mista no Estado do Paraná, Brasil Composition and dynamics of Alouatta guariba clamitans Cabrera, groups (Primates, Atelidae on an Araucaria Pine Forest remnant on Paraná State, Brazil

Directory of Open Access Journals (Sweden)

João M. D. Miranda

2005-03-01

Full Text Available As espécies do gênero Alouatta Lacépède, 1799 são sociais e formam grupos que mostram grandes variações na sua composição sexo-etária. Alouatta guariba clamitans Cabrera, 1940 é uma das espécies com escassos trabalhos que mostrem aspectos sociais deste primata. O estudo foi conduzido na Chácara Payquerê, localizada no Município de Balsa Nova, Estado do Paraná, Sul do Brasil. A área é inserida no contexto da Floresta Ombrófila Mista e faz parte da Área de Proteção Ambiental da Escarpa Devoniana. O trabalho foi conduzido de fevereiro de 2002 até junho de 2004, identificando grupos, registrando a composição sexo-etária e sua dinâmica. Foram identificados seis grupos na área de estudo, no decorrer dos três anos as médias de suas composições foram: 6,33 indivíduo/grupo; 1,47 Macho Adulto (MA/grupo; 0,46 Machos subadultos/grupo; 2,2 Fêmeas Adultas(FA/grupo; 1,13 Juvenis II/grupo; 0,8 Juvenis I/grupo e 0,27 Infantes/grupo. Para uma análise sobre a composição sexo-etária da espécie, os dados obtidos neste trabalho foram somados a 66 outros grupos de A. g. clamitans disponíveis na literatura (totalizando 72 grupos, sendo que as médias para a espécie foram: 6,25 indiv./grupo; 1,39 MA/grupo; 2,26 FA/grupo e 2,62 Indivíduos Imaturos/grupo. Quanto à dinâmica dos grupos registraram-se: 11 nascimentos; 25 mudanças de categoria etária; 4 indivíduos transeuntes; 1 morte/desaparecimento e a quebra de um grupo em dois. Este trabalho analisa a composição sexo-etária de A. g. clamitans além de discutir alguns aspectos da dinâmica dos grupos.Species in the genus Alouatta Lacépède, 1799 are social, and their groups display considerable variation in their age and sex composition. Little is known about the social organization of groups of Alouatta guariba clamitans Cabrera, 1940. This study was carried out between February 2002 and June 2004 in the Chácara Payquerê, located in the municipality of Balsa Nova in the

Grupo escolar de Alfarnate (Málaga: colegio y casas para maestros de José María Santos Rein

Directory of Open Access Journals (Sweden)

Sergio Romero Bueno

2015-01-01

Full Text Available Este artículo estudia y destaca el valor arquitectónico del grupo escolar de Alfarnate (un colegio y once viviendas para maestros, por su singularidad constructiva de claro tinte racionalista y por desarrollar un funcionalismo propio del Movimiento Moderno. Su atrevida modernidad, para un lugar alejado del núcleo urbano de Málaga, trascendió la arquitectura popular y tradicional para establecer su prevalencia como solución a los problemas de escolarización y de habitabilidad en la zona rural en los años sesenta del siglo XX. Documentamos y analizamos las vicisitudes que el arquitecto José María Santos Rein consiguió superar para realizar la obra y aportamos un estudio formal para dejar constancia de su existencia tras la inevitable demolición de las once casas en el año 2011.

Web didáctica: “Cabrera de Mar, arqueologia i patrimoni”. Un ejemplo de reconstrucción o anastilosis virtual de yacimientos y paisajes arqueológicos íberos, romanos y medievales. La infografía al servicio de la arqueología, del patrimonio y de la educación.

Directory of Open Access Journals (Sweden)

Josep Maria Rovira i Juan

2010-05-01

Full Text Available Presentation of the educational Web www.cabrerademarpatrimoni.cat “Cabrera de Mar, arqueología i patrimoni”. Web whose central objective disclosure (to the general public, but especially to our elementary and secondary students of archeology and heritage Iberian, Roman, medieval and modern, making it understandable to make the content and value that has no doubt to educate in the values of heritage and its preservation. The thread and the most characteristic element is anastilosis or virtual reconstruction of archaeological sites. Throughout the communication is intended to answer the question "what is, for what it is, and as for who is it? With the teaching’s Web that we present today.

Xylem anatomy of the Caesalpiniaceae registered in wood collection of the Universidad Distrital Francisco Jose de Caldas

International Nuclear Information System (INIS)

Pulido Rodriguez, Esperanza N; Mateus, Durley; Lozano D, Ivan

2011-01-01

The anatomical study of the xylem of 21 species of Caesalpiniaceae registered in the wood collection Jos Anatolio Lastra Rivera (JALR), of the Universidad Distrital Francisco Jose de Caldas, included the macroscopic, microscopic and biometric characterization of the cellular elements that constitute the xylematic tissue. These analyses were developed following parameters defined by the International Association Wood Anatomist Committee (IAWA Committee 1989) and methods established by the Wood Technology Laboratory of Universidad Distrital. Measurements and descriptions were used to prepare identification keys and similarity analysis. The wood of Caesalpiniaceae family illustrated common characteristics in growth rings differentiation, porosity, vessel arrangement, deposits, diameter and length, plates perforation type, alternate and vestured intervessel pits; fibers wall thickness and length; paratracheal axial vasicentric parenchyma, aliform and banded parenchyma and presence of prismatic crystals. Also, variations in anatomical features such us longitudinal channels were found as diagnostic for some genera like Copaifera. The variation and analysis of anatomical characteristics of the xylem tissue allowed to verify some taxonomic relations of the family Caesalpiniaceae, like the observed with the species Mora megistosperma, Mora oleofera, Peltogyne pubescens, Peltogyne paniculata, Sclerolobium odoratissimum and Tachigali polyphylla .

The meaning of nervios: a sociocultural analysis of symptom presentation in San Jose, Costa Rica.

Science.gov (United States)

Low, S M

1981-03-01

The foundation of the symbolic tradition in medical anthropology is the examination of a patient's experience of a category of illness. The interpretation of folk explanations of etiology and nosology provides insight into the cultural definition of what constitutes an illness, how and why an illness is labeled, and how the afflicted individual should be treated. Further, the analysis of sociocultural meaning emerges as a critical theoretical contribution to our understanding of health and culture. Allen Young in his article "Some Implications of Medical Beliefs and Practices for Social Anthropology" suggests " that if we want to learn the social meaning of sickness, we must understand that 'signs,' whatever their genesis, become 'symptoms' because they are expressed, elicited, and perceived in socially acquired ways" (1976: 14). He further states that some categories of sickness are particularly interesting in that they enable people to organize the illness event into an episode that has form and meaning (1976: 19-20). Nervios is an example of a symptom that has acquired a special sociocultural pattern of expression, elicitation and perception in San Jose, Costa Rica. The empirical study of symptom presentation in general medicine and psychiatric outpatient clinics describes the patients who present the symptom and their associated attributes and explanations of the symptom's occurrence. The meaning of nervios is then discussed within a social interactional and symbolic framework.

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

The Spanish Nuclear Safety Council and nuclear power stations in operation in Spain

International Nuclear Information System (INIS)

Perello, M.

1984-01-01

On 20 April 1980 the Spanish Congress of Deputies passed an Act setting up the Nuclear Safety Council (CSN) as the sole organization responsible for nuclear safety and radiation protection. In this paper it is stated that that date marked the beginning of a new nuclear safety policy in Spain. As one of its objectives, this policy is aimed at the monitoring and testing of operating nuclear installations. A detailed description is given of the Operating Nuclear Installation Service (SINE), including its basic structure, its functions and the technical and manpower resources available to it. The maintenance of close relations with other organs of the CSN is considered of paramount importance in order for the tasks allotted to SINE to be fulfilled. International co-operation and outside contracting greatly assist importing countries which have limited manpower resources. A description is then given of the present state of the nuclear power stations in operation in Spain together with an account of the most important initiatives which have been taken so far. The year 1968 saw the beginning of commercial operation of the Jose Cabrera nuclear power station, which has the only single-loop PWR reactor in the world. At present, it is being subjected to the Systematic Evaluation Programme (SEP). The Santa Maria de Garona nuclear power station has been operating for over twelve years and is also being subjected to the SEP although design modifications derived from operating experience have already been introduced. The Vandellos I station was the last of the first generation and has also benefited from the operating experience of similar French plants. Unit 1 of the Almaraz power station opens the door to the second generation and the generic problem which has occurred with the steam generators is in process of being solved. Lastly, some general conclusions are presented about the organization of and experience acquired with operating nuclear power stations. (author)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Revisión del género PerezIa (compositae

Directory of Open Access Journals (Sweden)

Liliana Katinas

2012-06-01

Full Text Available El género Perezia Lag. (Compositae se compone de 30 especies de hierbas perennes distribuidas principalmente en la región andino-patagónica, desde Colombia hasta el sur de Argentina y Chile, con algunas especies en el este de Brasil y de la Argentina, Paraguay y Uruguay. Sus corolas bilabiadas, anteras sagitadas, estilos truncados con una coronita apical de pelos colectores y su estructura polínica ubican al género en la tribu Nassauvieae, subfamilia Mutisioideae. Se propone en este trabajo: a la rehabilitación de Perezia atacamensis, P. burkartii y P. capito; b la transferencia de P. ciliosa var. dentata a la categoría de especie [P. dentata (Cabrera comb. nov.]; c la inclusión de Perezia catharinensis Cabrera, P. eryngioides (Cabrera Crisci y P. volcanensis Cabrera, descritas o transferidas posteriormente a la primer revisión del género en 1969; d la eliminación de subespecies en P. lactucoides; e el establecimiento de la sinonimia entre P. pinnatifida, P. pygmaea y P. coerulescens; f el establecimiento de la sinonimia entre P. pungens, P. ciliaris, P. mandonii, P. purpurata, P. carduncelloides y P. sublyrata; g la rehabilitación de Perezia recurvata subsp. recurvata, subsp. beckii y subsp. triceps. Se designaron dos lectotipos y cinco neotipos, y se excluyeron 86 nombres del género.Revision of the genus Perezia (Compositae. The genus Perezia Lag. (Compositae, with 30 species, is distributed in the Andean-Patagonian region from Colombia to southern Argentina and Chile, and in eastern Brasil and Argentina, Paraguay and Uruguay. It belongs to the tribe Nassauvieae, subfamily Mutisioideae by its bilabiate corollas, caudate anthers, truncate style branches with a crown of collector hairs, and a particular pollen structure. It is proposed in this paper: a the rehabilitation of Perezia atacamensis, P. burkartii, and P. capito; b the change of P. ciliosa var. dentata to the species category [P. dentata (Cabrera comb. nov.]; c the

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Isotopic investigation of ground water resources in the Ojo Alamo sandstone, Nacimiento, and San Jose Formations, San Juan Basin, New Mexico. Technical completion report

International Nuclear Information System (INIS)

Phillips, F.M.; Peeters, L.A.; Tansey, M.K.

1984-06-01

The San Juan Basin, in northwest New Mexico, has vast reserves of strippable, low-sulfur coal. Development of the resource will require large quantities of water, from an area where water resources are not abundant. Since surface-water supplies are fully allocated, increased future water demands will have to be met through ground-water development. The study concentrates on the Ojo Alamo, Nacimiento, and San Jose Formations, the aquifers directly above the principal coal unit. Carbon-14 and tritium methods were used to date the ground water in these units. Initial radiocarbon activities were calculated using the models of Vogel, Tamers, Pearson, Mook and Fontes. The observation lends support to the hypothesis of isotopically lighter Pleistocene precipitation. Such lighter recharge was most likely due to a colder mean annual temperature and perhaps increased winter precipitation. A similar change is obtained from noble-gas paleothermometry

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

Current situation of sexual and reproductive health of men deprived of liberty in the Institutional Care Center of San Jose

Directory of Open Access Journals (Sweden)

Dorita Rivas Fonseca

2013-10-01

Full Text Available The objective of this research was to determine the current status of the issue of sexual and reproductive health ofthe prisoners Institutional Care Center (CAI of San Jose. It is a descriptive study. Through a strategic samplingdetermined the participation of 102 men. The information was obtained by applying a self-administeredquestionnaire with closed and open questions. As a result relevant to your socio-demographic profile, it appearsthat deprived of their liberty is a very heterogeneous group. As regards sexual and reproductive health, the firstconcept they relate to the prevention of disease and the second reproductive aspects, this shows limitations inknowledge on the topics, something that affects the daily life activities and self-care. It is concluded that researchby nurses Gyneco-obstetric in the deprived of liberty is almost null not only in the country but in the world,especially if it comes with the male population. In the case of CAI Prison, health care is not enough for thenumber of inmates who inhabit (overpopulation of almost 50%, this implies a deterioration in health and physicalcondition of these people, as well as sexual and reproductive health

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Spanish RPV head penetrations. Regulatory status

International Nuclear Information System (INIS)

Figueras, J.M.; Colino, J.R.

1995-01-01

The paper presents the actual status of inspection results on the Spanish PWR RPV CRD head penetrations (CRDH's), after two years of a whole program of inspections in all affected plants. Actual situation of penetrations pertaining to ALMARAZ 1 and 2, ASCO 1 and 2 and VANDELLOS 2 NPP's show any damage in those CRDH's inspected in 1993 and 1994 (roughly 20 out of 65 CRDH's at each unit). The paper presents a summary of CRDH characteristics, inspection methods and results obtained in each plant. TRILLO NPP has a different CRDH design (KWU-SIEMENS type) and for that reason is not considered an affected plant nor has conducted any inspection up to now. JOSE CABRERA (ZORITA) NPP has shown extensive damage, both in the lower side (weldment to the vessel) and in the upper free span area, near bimetallic weldment to SS 304, in active and nonactive penetrations and also in the vent nozzle. The paper comments extensively on the CRDH materials general data, root-cause analysis and structural analysis of degraded zones, inspection results, repair actions and other additional actions applied up to now. Finally, the paper deals with the regulatory actions taken by CSN on this topic, both for those NPP's actually non affected by the IGSCC phenomenon in the RPV CRDH's and for the specific safety case of ZORITA NPP. (author)

DOSEFU: Computer application for dose calculation and effluent management in normal operation

International Nuclear Information System (INIS)

Martin Garcia, J. E.; Gonzalvo Manovel, A.; Revuelta Garcia, L.

2002-01-01

DOSEFU is a computer application on Windows that develops the methodology of nuclear power plant Exterior Dose Calculation Manuals (Manuals de Calculo de Dosis al Exterior-MACADE) for calculating doses in normal operation caused by radioactive liquid and gaseous effluents, for the purpose of enforcing the new Spanish Regulation on Health Protection against Ionizing Radiations, Royal Decree 783/2001 resulting from transposition of Directive 96/29/Euratom whereby the basic rules regarding health protection of workers and the population against risks resulting from ionizing radiations are established. In addition to making dose calculations, DOSEFU generates, on a magnetic support, the information regarding radioactive liquid and gaseous effluents that plants must periodically send to the CSN (ELGA format). The computer application has been developed for the specific case of Jose Cabrera NPP, which is called DOEZOR. This application can be easily implemented in any other nuclear or radioactive facility. The application is user-friendly, as the end user inputs data and executes the different modules through keys and dialogue boxes that are enabled by clicking on the mouse (see figures 2, 3, 4 and 5 ), The application runs under Windows 95. Digital Visual Fortran has been used as the development program, as this does not require additional libraries (DLLs), it can be installed in any computer without affecting other programs that are already installed. (Author)

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

SRTM Perspective View with Landsat Overlay: San Jose, Costa Rica

Science.gov (United States)

2001-01-01

This perspective view shows the capital city of San Jose, Costa Rica, in the right center of the image (gray area). Rising behind it are the volcanoes Irazu, 3402 meters high (11,161 feet) and Turrialba, 3330 meters high (10,925 feet.)Irazu is the highest volcano in Costa Rica and is located in the Irazu Volcano National Park, established in 1955. There have been at least 23 eruptions of Irazu since 1723, the most recent during 1963 to 1965. This activity sent tephra and secondary mudflows into cultivated areas, caused at least 40 deaths, and destroyed 400 houses and some factories.This image was generated in support of the Central American Commission for Environment and Development through an agreement with NASA. The Commission involves eight nations working to develop the Mesoamerican Biological Corridor, an effort to study and preserve some of the most biologically diverse regions of the planet.This three-dimensional perspective view was generated using topographic data from the Shuttle Radar Topography Mission (SRTM) and an enhanced false-color Landsat 7 satellite image. Colors are from Landsat bands 5, 4, and 2 as red, green and blue, respectively. Topographic expression is exaggerated 2X.Landsat has been providing visible and infrared views of the Earth since 1972. SRTM elevation data matches the 30-meter resolution of most Landsat images and will substantially help in analyses of the large and growing Landsat image archive. The Landsat 7 Thematic Mapper image used here was provided to the SRTM by the United States Geological Survey, Earth Resources Observation Systems (EROS) Data Center, Sioux Falls, South Dakota.Elevation data used in this image was acquired by the Shuttle Radar Topography Mission (SRTM) aboard the Space Shuttle Endeavour, launched on February 11,2000. SRTM used the same radar instrument that comprised the Spaceborne Imaging Radar-C/X-Band Synthetic Aperture Radar (SIR-C/X-SAR) that flew twice on the Space Shuttle Endeavour in 1994. SRTM

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Reactor physics and reactor computations

International Nuclear Information System (INIS)

Ronen, Y.; Elias, E.

1994-01-01

Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Reactor

International Nuclear Information System (INIS)

Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

1979-01-01

Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Reactor container

International Nuclear Information System (INIS)

Kato, Masami; Nishio, Masahide.

1987-01-01

Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Evaluation of heavy metals (Cr, Fe, Ni, Cu, Zn, Cd, Pb and Hg) in water, sediments and water lily (Eichornia crassipes) from Jose Antonio Alzate dam; Evaluacion de metales pesados Cr, Fe, Ni, Cu, Zn, Cd, Pb y Hg en agua, sedimento y lirio acuatico (Eichhornia crassipes) de la Presa Jose Antonio Alzate, Estado de Mexico

Energy Technology Data Exchange (ETDEWEB)

Avila P, P

1996-12-31

Water, sediments and water lily (Eichornia crassipes) from the Jose Antonio Alzate Dam were analyzed in order to determine concentrations of chromium, iron, nickel, copper, zinc, cadmium, lead and mercury. Mercury, lead, chromium and iron were found in concentrations above permissible limits in water, and in high concentrations in sediments. Cadmium, nickel, copper and zinc never were found in concentrations above permissible limits in water. The highest concentrations of heavy metals in water lily were found in the root. Accumulation factors decreased in the following order: Zn> Cr> Fe> Ni> Cu> Pb> Hg and Cd. Statistical differences ({alpha} < 0.5) between the collection samples dates was observed. High correlations between metals concentrations in superficial water, sediment and water hyacinth were also observed. These correlations could indicate that the heavy metals studied here, are originated from a natural source such as sediments or from an industrial source. (Author).

VASCULAR EPIPHYTE COMPONENT OCCURING IN URBAN TREES IN THE SQUARE PROFESSOR JOSE INACIO IN PIRATININGA COUNTY, SÃO PAULO STATE

Directory of Open Access Journals (Sweden)

Juliano Ricardo Fabricante

2006-12-01

Full Text Available Epiphytes are plants cropped in other plants to damage them. The main characteristics of this group of plants is unknown in this part of Brazil, making it important to study this subject. The aim of this work was to characterize the structural and the floristic composition of vascular epiphytes on urban trees; to classify ecological categories species, according to hoist and to calculate community diversity. The studied area (Professor Jose Inacio Square, standed at the municipal district of Piratininga, Sao Paulo State. The climate is Cwa according to Köeppen classification and sandy phase dark red oxisol (latosol. It had been identified 10 species encompassing 6 gender and 6 families, being with best improvement the Bromeliaceae, representing 30 % of data. Tillandsia, with 3 species, was the most abundant gender and featured holoepiphytes were represented, with 70% species. About epiphyte community structure: Tillandsia recurvata and Tillandsia tricholepis were the most important (with bigger VIe in the studies community. Diversity was 2.371 nats/individual to specie and 2.04 nats/individual per family. The small number of species and low diversity index should be due to local conditions, urban area, with different features from that founded at normal eco-systems.

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Evaluation of the levels of particles PM{sub 10} and nitrogen dioxide at the city of San Jose, Costa Rica: 2005-2006; Evaluacion de los niveles de particulas PM{sub 10} y dioxido de nitrogeno en la ciudad de San Jose, Costa Rica: 2005-2006

Energy Technology Data Exchange (ETDEWEB)

Herrera-Murillo, Jorge [Univ. Nacional, Escuela de Ciencias Ambientales, Lab. de Analisis Ambiental, Campus Omar Dengo, apartado 86-3000 Heredia (Costa Rica). Tel: 506-277-3292; Rodriguez Roman, Susana, E-mail: rr.susana@gmail.com, E-mail: susanar@cariari.ucr.ac.cr [Univ. de Costa Rica, Escuela de Quimica, Ciudad Universitaria Rodrigo Facio, San Pedro de Montes de Oca (Costa Rica). Tel: 506-2207-5376

2009-07-15

The levels of particulate matter PM{sub 10} were determined at two sites of San Jose City (Catedral Metropolitana and Junta de Educacion), during a year (September 2005-September 2006), obtaining as annual average 36 {+-} 8 {mu}g/m{sup 3} and 25 {+-} 7 {mu}g/m{sup 3}, respectively. Also, the levels of sulfates, nitrates and chlorides were measured for both sites. The annual averages have resulted not to be significantly different for both sites, with a level of significance of 5%. Three of the fourteen sites of measurement of the concentration of nitrogen dioxide in the city, have presented higher values to the recommendation of the World Health Organization, monitoring for one month. The principal component analysis that were applied to the data of this gas, has showed that the variations in the levels are due to large-scale phenomena (meteorological). However, the concentration of sulfate present in the particulate matter has reached higher values to those recorded in cities like Rio de Janeiro and Mexico D.F., product probably of the high sulfur content present in the fuels that are used by the vehicle fleet at Costa Rica. The principal component analysis indicates that has existed a strong correlation between the concentrations of sulfate and nitrate present in the particles PM{sub 10}, pointing the contribution of anthropogenic sources. In the case of chloride, has highlighted the strong existing relationship with the meteorological parameters that were registered during the sampling period. (author) [Spanish] Los niveles de material particulado PM{sub 10} fueron determinados en dos sitios de la ciudad de San Jose (Catedral Metropolitana y Junta de Educacion), durante un ano (setiembre 2005-setiembre 2006), obteniendo como promedio anual 36 {+-} 8 {mu}g/m{sup 3} y 25 {+-} 7 {mu}g/m{sup 3}, respectivamente. Ademas, los niveles de sulfatos, nitratos y cloruros fueron medidos para ambos sitios. Las medias anuales han resultado no ser significativamente diferentes

Slurry reactors

Energy Technology Data Exchange (ETDEWEB)

Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

1979-08-01

Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

Evaluation of heavy metals (Cr, Fe, Ni, Cu, Zn, Cd, Pb and Hg) in water, sediments and water lily (Eichornia crassipes) from Jose Antonio Alzate dam

International Nuclear Information System (INIS)

Avila P, P.

1995-01-01

Water, sediments and water lily (Eichornia crassipes) from the Jose Antonio Alzate Dam were analyzed in order to determine concentrations of chromium, iron, nickel, copper, zinc, cadmium, lead and mercury. Mercury, lead, chromium and iron were found in concentrations above permissible limits in water, and in high concentrations in sediments. Cadmium, nickel, copper and zinc never were found in concentrations above permissible limits in water. The highest concentrations of heavy metals in water lily were found in the root. Accumulation factors decreased in the following order: Zn> Cr> Fe> Ni> Cu> Pb> Hg and Cd. Statistical differences (α < 0.5) between the collection samples dates was observed. High correlations between metals concentrations in superficial water, sediment and water hyacinth were also observed. These correlations could indicate that the heavy metals studied here, are originated from a natural source such as sediments or from an industrial source. (Author)

Three Newly Naturalized Plants in Taiwan

Directory of Open Access Journals (Sweden)

Ming-Jer Jung

2009-11-01

Full Text Available Three newly naturalized plants are reported in this paper. Hypochaeris microcephala (Sch. Bip. Cabrera var. albiflora (Kuntze Cabrera (Asteraceae is naturalized in urban areas of northern Taiwan. Indigofera pseudo-tinctoria Matsum. (Leguminosae is naturalized in low elevations of northern and southern Taiwan and in middle elevations of central Taiwan. Lamium purpureum L. (Laminaceae has become naturalized locally in middle elevations of central Taiwan. Descriptions, illustrations and color photos of these plants are provided.

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Space-time reactor kinetics for heterogeneous reactor structure

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

1969-11-15

An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Feasible reactor power cutback logic development for an integral reactor

International Nuclear Information System (INIS)

Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

2013-01-01

Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

RA Reactor

International Nuclear Information System (INIS)

1978-02-01

In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

H Reactor

Data.gov (United States)

Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

Heavy metal contents in sediments of rivers and adjacent streams to the filling of Rio Azul, Rio Azul, San Jose, Costa Rica

International Nuclear Information System (INIS)

Mora Amador, Raul Alberto

2003-01-01

In Costa Rica there are no studies that have documented the concentrations of heavy metals in river sediments, caused by point sources of pollutants, such as the deposits sites of solid waste. The potential source of heavy metals concentrations in the study area, is an active landfill, known as Relleno Sanitario de Rio Azul, which is located southeast of the city of San Jose, in Distrito de Rio Azul de La Union. The site where the landfill Rio Azul is located is composed by a sequence of volcanic materials, which could have been affected by local faulting. This fact would mean a serious threat of contamination of groundwater in the region, geological faults are therefore a highly permeable, able to get in touch the fill pollutants with some aquifer. The mass of metal deposited on the site, from 1978 to date, has been estimated between 70000-100000 tm, which it becomes environmental concern in a first order. The results of X-ray fluorescence show the presence of elements such as vanadium, chromium, magnesium, copper, zinc, rubidium, strontium, potassium, calcium, iron, titanium and yttrium, which show a trend towards more high Concentrates in the fine fraction (d [es

Technical proposal on the treatment of the influent of the Jose Antonio Alzate dam using aerobic pond system; Propuesta tecnica sobre el tratamiento del influente de la presa Jose Antonio Alzate a traves un sistema lagunar aerobio

Energy Technology Data Exchange (ETDEWEB)

Rosas de Alba, S.G. [Instituto Tecnologico de Toluca, Metepec, Estado de Mexico (Mexico)]. E-mail: sergio.rosas01@cfe.gob.mx; Barcelo Quintal, I.D.; Bussy Beaurain, A.L.; lopez Galvan, E. [Universidad Autonoma Metropolitana, Unidad Azcapotzalco (Mexico)]. E-mail: idbq@correo.azc.uam.mx; leemp@correo.azc.uam.mx; loge@correo.azc.uam.mx

2010-01-15

In Mexico, discharge of waste water not treated is a common problem; this is the case of the Lerma river, where the original clean water contribution of the Almoloya-Lerma-San Bartolo timber-roof system, located in the valley of Toluca, State of Mexico, was replaced by the wastewater of the valley. The attempts to clean the water before their spill have been unfruitful and this has caused the present anaerobic conditions of the river in its entire route including the first receiving stage that is the Jose Antonio Alzate Dam (PJAA), becoming a potential source of diseases and a useless water body. In this work we evaluate the treatment efficiency of experimental waste water pond system. The treatment system was constructed with 5 screens which formed 6 internal channels and it had an operation in continuous form by 26 weeks, providing it with continuously polluted water of the Lerma river. We found that the efficiency of reduction of the organic matter reported as BOD, was 83% and it is carried out in the first channel after 7 days of hydraulic retention verifying that the channels do not work like lagoons in series. With the collected data we calculated, through the application of design equations, the factor of decay of organic matter k and the number of dispersion d, specific for the type of treatment and environmental conditions of the site. The required surface of treatment calculated to reach the efficiency of reduction of the DOD5 obtained experimentally. Thus, a surface of: 1 281.6 hectares distributed in 13 lagoons that can deal with a volume 8646.70 m{sup 3}/h was determined. Since this was a bigger surface than that of the dam, the available area was evaluated to make a proposal closer to reallity. [Spanish] La descarga de aguas residuales a los sistemas acuaticos, sin previo tratamiento, es un problema generalizado en Mexico; como es el caso del rio Lerma en donde se sustituyo el aporte de agua limpia proveniente del sistema lagunar Almoloya

Reactor container

International Nuclear Information System (INIS)

Naruse, Yoshihiro.

1990-01-01

The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

Hybrid reactors

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

Multiregion reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

Reactor building

International Nuclear Information System (INIS)

Maruyama, Toru; Murata, Ritsuko.

1996-01-01

In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Measuring and building resilience after big storms: Lessons learned from Super-Storm Sandy for the Harvey, Irma, Jose, and Maria coasts

Science.gov (United States)

Murdoch, P. S.; Penn, K. M.; Taylor, S. M.; Subramanian, B.; Bennett, R.

2017-12-01

As we recover from recent large storms, we need information to support increased environmental and socio-economic resilience of the Nation's coasts. Defining baseline conditions, tracking effects of mitigation actions, and measuring the uncertainty of resilience to future disturbance are essential so that the best management practices can be determined. The US Dept. of the Interior invested over $787 million dollars in 2013 to understand and mitigate coastal storm vulnerabilities and enhance resilience of the Northeast coast following Super-Storm Sandy. Several lessons-learned from that investment have direct application to mitigation and restoration needs following Hurricanes Harvey, Irma, Jose and Maria. New models of inundation, overwash, and erosion, developed during the Sandy projects have already been applied to coastlines before and after these recent storms. Results from wetland, beach, back-bay, estuary, and built-environment projects improved models of inundation and erosion from surge and waves. Tests of nature-based infrastructure for mitigating coastal disturbance yielded new concepts for best-practices. Ecological and socio-economic measurements established for detecting disturbance and tracking recovery provide baseline data critical to early detection of vulnerabilities. The Sandy lessons and preliminary applications on the recent storms could help define best-resilience practices before more costly mitigation or restoration efforts are required.

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

International Nuclear Information System (INIS)

Eustache, S.

1999-01-01

This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Nuclear reactor instrumentation at research reactor renewal

International Nuclear Information System (INIS)

Baers, B.; Pellionisz, P.

1981-10-01

The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

BWR type reactor

International Nuclear Information System (INIS)

Watanabe, Shoichi

1983-01-01

Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

Reactor core of nuclear reactor

International Nuclear Information System (INIS)

Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

1994-01-01

In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

Theater and the Discourse on Power: Jose Rizal’s Participation in Philippine Theater in the Last Decades of the Nineteenth Century

Directory of Open Access Journals (Sweden)

Apolonio B. Chua

2007-06-01

Full Text Available The study focuses on Jose Rizal’s participation in Philippine Theater during the last decades of the nineteenth century. It starts with a careful inventory of attitudes towards existing theater forms and a description of the culture of theater as conceived and imagined in Rizal’s Noli Me Tangere (1887 and El Filibusterismo (1891. In Chapter 20: “The Town Council’s Meeting” in Noli Me Tangere, the study zooms in on the debate on what would be the best and most appropriate theatre piece for the town fiesta. Here, Rizal delineates theater enmeshed in issues of power and Spanish colonialism. He notes as significant the ilustrados’ claim to theatre space and ideology, thereby interrogating Spanish hegemony. The debate becomes the central imagery and situation for Rizal’s analysis and construction of the history of Philippine theater in the novel. Conservative and radical elements duel. The conflict becomes sharper as Rizal continues his critique by putting into his fictional world the very historical actors known at that time; namely, Nemesio Ratia, Jose Carvajal and Praxedes Julia Fernandez (also known as “Yeyeng”. They were part of the comedia troupe hired by the town for the fiesta. In Rizal’s second novel, El Filibusterismo, we encounter the events surrounding the presentation by a French opera troupe in Teatro de Variedades, which Rizal considers as the Manila theater model. The features of this model include a particular ticket system, various kinds of audiences, imported dramatic texts which were largely incomprehensible, actors behaving as actors both on-stage and off-stage, and the Teatro de Variedades space as stage for seizure or possession of power. When the students in the audience stage a walk-out in the theatre of the city and when in provincial San Diego, a stampede cuts short a comedia performance, the interrelationships between society and a discourse of power are revealed.Rizal’s annotations of Philippine

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

Heterogeneous reactors

International Nuclear Information System (INIS)

Moura Neto, C. de; Nair, R.P.K.

1979-08-01

The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

Nuclear reactor neutron shielding

Science.gov (United States)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

2017-09-12

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

Fast breeder reactors

International Nuclear Information System (INIS)

Heinzel, V.

1975-01-01

The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

Reactor container

International Nuclear Information System (INIS)

Shibata, Satoru; Kawashima, Hiroaki

1984-01-01

Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

Reactor container

International Nuclear Information System (INIS)

Fukazawa, Masanori.

1991-01-01

A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

A multi-purpose reactor

International Nuclear Information System (INIS)

Changwen Ma

2000-01-01

An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

Observations of basin ground motions from a dense seismic array in San Jose, California

Science.gov (United States)

Frankel, A.; Carver, D.; Cranswick, E.; Bice, T.; Sell, R.; Hanson, S.

2001-01-01

We installed a dense array of 41 digital seismographs in San Jose, California, to evaluate in detail the effects of a deep sedimentary basin and shallow sedimentary deposits on earthquake ground motions. This urban array is located near the eastern edge of the Santa Clara Valley and spans the Evergreen sedimentary basin identified by gravity data. Average station spacing is 1 km, with three stations initially spaced 110 m apart. Despite the high-noise urban environment, the stations of the array successfully triggered on and recorded small local earthquakes (M 2.5-2.8 at 10-25 km distance) and larger regional events such as the M 5.0 Bolinas earthquake (90 km distance), M 4.6-5.6 earthquakes near Mammoth Lakes (270 km distance), M 4.9-5.6 events in western Nevada (420 km distance) and the M 7.1 Hector Mine earthquake (590 km distance). Maps of spectral ratios across the array show that the highest amplitudes in all frequency bands studied (0.125-8 Hz) are generally observed at stations farther from the eastern edge of the Santa Clara Valley. Larger spectral amplitudes are often observed above the western edge of the Evergreen Basin. Snapshots of the recorded wavefield crossing the array for regional events to the east reveal that large, low-frequency (0.125-0.5 Hz) arrivals after the S-wave travel from south to north across the array. A moving-window, cross-correlation analysis finds that these later arrivals are surface waves traveling from the south. The timing and propagation direction of these arrivals indicates that they were likely produced by scattering of incident S waves at the border of the Santa Clara Valley to the south of the array. It is remarkable that the largest low-frequency phases at many of the valley sites for regional events to the east are basin surface waves coming from a direction about 70 degrees different from that of the epicenters. Basin surface waves emanating from the eastern edge of the valley are also identified by the cross

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Supercritical Water Reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

2014-01-01

The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

New about research reactors

International Nuclear Information System (INIS)

Egorenkov, P.M.

2001-01-01

The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

International Nuclear Information System (INIS)

Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

2013-01-01

The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

Design guide for Category III reactors: pool type reactors

International Nuclear Information System (INIS)

Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

1978-11-01

The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

The computerized reactor period measurement system for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1996-01-01

The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

Computer measurement system of reactor period for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1997-01-01

The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

Reactor safety method

International Nuclear Information System (INIS)

Vachon, L.J.

1980-01-01

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

1989-01-01

In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

Physics of nuclear reactors

International Nuclear Information System (INIS)

Baeten, Peter

2006-01-01

This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

Guide to power reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1959-07-15

The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

Analysis of dynamic stability and safety of reactor system by reactor simulator

International Nuclear Information System (INIS)

Raisic, N.

1963-11-01

In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

Nuclear reactor physics

CERN Document Server

Stacey, Weston M

2010-01-01

Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

Atomic reactor thermal engineering

International Nuclear Information System (INIS)

Kim, Gwang Ryong

1983-02-01

This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

Argentinean integrated small reactor design and scale economy analysis of integrated reactor

International Nuclear Information System (INIS)

Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

2000-01-01

This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

Fast reactors

International Nuclear Information System (INIS)

Vasile, A.

2001-01-01

Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

Reactor power measuring device

International Nuclear Information System (INIS)

Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

1993-01-01

The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

Reactor container

International Nuclear Information System (INIS)

Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

1997-01-01

In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

To the analysis of reactor noise in boiling water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1972-01-01

The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

State space modeling of reactor core in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

2014-07-10

The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

Virtual nuclear reactor for education of nuclear reactor physics

International Nuclear Information System (INIS)

Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

2008-01-01

As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

International Nuclear Information System (INIS)

Schikorr, W.M.

In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

Reactor core

International Nuclear Information System (INIS)

Azekura, Kazuo; Kurihara, Kunitoshi.

1992-01-01

In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

Nuclear reactor

International Nuclear Information System (INIS)

Batheja, P.; Huber, R.; Rau, P.

1985-01-01

Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

Energy Technology Data Exchange (ETDEWEB)

Kliem, S.

2017-09-15

The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

The fast breeder reactor

International Nuclear Information System (INIS)

Collier, J.

1990-01-01

The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Ulysse, mentor reactor

International Nuclear Information System (INIS)

Bouquin, B.; Rio, I.; Safieh, J.

1997-01-01

On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

Science.gov (United States)

Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

2016-01-01

Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

The fast reactor

International Nuclear Information System (INIS)

1980-02-01

The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Masood, Z.

2016-01-01

The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

Fast Spectrum Reactors

CERN Document Server

Todd, Donald; Tsvetkov, Pavel

2012-01-01

Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1985-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

1982-09-01

Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

International Nuclear Information System (INIS)

D. Kokkinos

2005-01-01

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

International Nuclear Information System (INIS)

Shiba, T.; Fallot, M.

2015-01-01

To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

Hierarchical spatial segregation of two Mediterranean vole species: the role of patch-network structure and matrix composition.

Science.gov (United States)

Pita, Ricardo; Lambin, Xavier; Mira, António; Beja, Pedro

2016-09-01

According to ecological theory, the coexistence of competitors in patchy environments may be facilitated by hierarchical spatial segregation along axes of environmental variation, but empirical evidence is limited. Cabrera and water voles show a metapopulation-like structure in Mediterranean farmland, where they are known to segregate along space, habitat, and time axes within habitat patches. Here, we assess whether segregation also occurs among and within landscapes, and how this is influenced by patch-network and matrix composition. We surveyed 75 landscapes, each covering 78 ha, where we mapped all habitat patches potentially suitable for Cabrera and water voles, and the area effectively occupied by each species (extent of occupancy). The relatively large water vole tended to be the sole occupant of landscapes with high habitat amount but relatively low patch density (i.e., with a few large patches), and with a predominantly agricultural matrix, whereas landscapes with high patch density (i.e., many small patches) and low agricultural cover, tended to be occupied exclusively by the small Cabrera vole. The two species tended to co-occur in landscapes with intermediate patch-network and matrix characteristics, though their extents of occurrence were negatively correlated after controlling for environmental effects. In combination with our previous studies on the Cabrera-water vole system, these findings illustrated empirically the occurrence of hierarchical spatial segregation, ranging from within-patches to among-landscapes. Overall, our study suggests that recognizing the hierarchical nature of spatial segregation patterns and their major environmental drivers should enhance our understanding of species coexistence in patchy environments.

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

NUCLEAR REACTOR

Science.gov (United States)

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

Reactor control device

International Nuclear Information System (INIS)

Fukami, Haruo; Morimoto, Yoshinori.

1981-01-01

Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

Reactor safety device

International Nuclear Information System (INIS)

Okada, Yasumasa.

1987-01-01

Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1980-09-01

Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

Reactor scram device for FBR type reactor

International Nuclear Information System (INIS)

Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

1991-01-01

In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Refuelling nuclear reactors

International Nuclear Information System (INIS)

Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

1981-01-01

An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

Compilation of reactor physics data of the year 1984, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-12-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Compilation of reactor physics data of the year 1983, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-06-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Reactor containment

International Nuclear Information System (INIS)

Kawabe, Ryuhei; Yamaki, Rika.

1990-01-01

A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1978-10-01

Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

NARCIS (Netherlands)

Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

2009-01-01

In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

Undergraduate reactor control experiment

International Nuclear Information System (INIS)

Edwards, R.M.; Power, M.A.; Bryan, M.

1992-01-01

A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

Snoj, L.; Kromar, M.; Zerovnik, G.

2012-01-01

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Reactor feedwater system

International Nuclear Information System (INIS)

Hikabe, Katsumi.

1978-01-01

Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

New reactor concepts

International Nuclear Information System (INIS)

Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

1998-11-01

The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Fission reactors and materials

International Nuclear Information System (INIS)

Frost, B.R.T.

1981-12-01

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J. Jr.

1981-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

Method of operating a reactor

International Nuclear Information System (INIS)

Oosumi, Katsumi; Yamamoto, Michiyoshi.

1980-01-01

Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

Reactor container

International Nuclear Information System (INIS)

Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

1987-01-01

Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1993-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor engineering department annual report

International Nuclear Information System (INIS)

1990-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

1977-09-01

Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1984-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1976-09-01

Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

Department of reactor technology

International Nuclear Information System (INIS)

1980-01-01

The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Reactor Safety Commission Code of Practice for Pressurized Water Reactors

International Nuclear Information System (INIS)

1990-01-01

The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

Remote Reactor Monitoring

Energy Technology Data Exchange (ETDEWEB)

Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-10-21

The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

2001-07-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

Canada-India Reactor (CIR)

Energy Technology Data Exchange (ETDEWEB)

None

1960-12-15

Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

International Nuclear Information System (INIS)

Nelson, W.R.

1982-01-01

REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

One piece reactor removal

International Nuclear Information System (INIS)

Chia, Wei-Min; Wang, Song-Feng

1993-01-01

The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

Nuclear reactors

International Nuclear Information System (INIS)

Middleton, J.E.

1977-01-01

Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

Reactor decommissioning

International Nuclear Information System (INIS)

Lawton, H.

1984-01-01

A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

2000-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1998-01-01

Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Cupac, S.; Vukadin, Z.

1996-01-01

Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

FBR type reactor core

International Nuclear Information System (INIS)

Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

1995-01-01

Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

Reactor operation method

International Nuclear Information System (INIS)

Osumi, Katsumi; Miki, Minoru.

1979-01-01

Purpose: To prevent stress corrosion cracks by decreasing the dissolved oxygen and hydrogen peroxide concentrations in the coolants within a reactor container upon transient operation such as at the start-up or shutdown of bwr type reactors. Method: After a condensate has been evacuated, deaeration operation is conducted while opening a main steam drain line, as well as a main steam separation valve and a by-pass valve in a turbine by-pass line connecting the main steam line and the condenser without by way of a turbine, and the reactor is started-up by the extraction of control rods after the concentration of dissolved oxygen in the cooling water within a pressure vessel has been decreased below a predetermined value. Nuclear heating is started after the reactor water has been increased to about 150 0 C by pump heating after the end of the deaeration operation for preventing the concentration of hydrogen peroxide and oxygen in the reactor water from temporarily increasing immediately after the start-up. The corrosive atmosphere in the reactor vessel can thus be moderated. (Horiuchi, T.)

Nuclear reactor buildings

International Nuclear Information System (INIS)

Nagashima, Shoji; Kato, Ryoichi.

1985-01-01

Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

The reactor Cabri

International Nuclear Information System (INIS)

Ailloud, J.; Millot, J.P.

1964-01-01

It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m 3 /h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

Dosimetry of radium equivalent in construction material of brick works in Sao Jose do Sabugi City - Paraiba, Brazil

International Nuclear Information System (INIS)

Araujo, Eduardo Eudes Nobrega de; Santos Junior, Jose Araujo dos; Amaral, Romilton dos Santos; Santos, Josineide Marques do Nascimento; Spacov, Isabel Cristina Guerra; Fernandez, Zahily Herrero

2015-01-01

The earth's crust has in its composition the Naturally Occurring Radioactive Material (NORM) that may have increased concentration due to activities of exploration and extraction of environmental resources. The civil construction is an economic activity that requires the use of much of the natural resources, such as the raw material of brick works, like clays, mainly used for the production of bricks and tiles. These construction materials may contain high levels of natural radioactive elements, even with concentrations higher than the limits established, given that the levels vary according to the composition of rocks and soil, due to the geological formation and may result in increased exposure of humans to natural radioactive activities. In this context, the radioecological dosimetry is defined in terms of Radium Equivalent activity (Ra eq ), that ensure radiometric conditions for the use of material derived from clays before its final application in housing construction, an initiative that ensures the radioecological safety of population. Thus, this study aimed to establish the calculation of Ra eq in the raw material of brick works located in Sao Jose do Sabugi city, state of Paraiba, in an area adjacent to the uranium deposits of Espinharas, to estimate the risks associated with primordial radionuclides attributed to TENORM activities (Technologically Enhanced Naturally Occurring Radioactive Materials) from the extraction and use of clay as a raw material in the manufacture of bricks and tiles. Analyses were performed by High Resolution Gamma Spectrometry, with HPGe-Be detector, assuming the state of secular radioactive equilibrium. The results ranged from 183.2 to 747.78 Bq/kg, with an average of 494.6 Bq/kg which exceeded the limit of 370 Bq/kg established by UNSCEAR for construction materials. Some samples obtained values exceeded by up the double this limit, suggesting control and radiometric certification for application of this material. (author)

An Account of the Accessioned Specimens in the Jose Vera Santos Memorial Herbarium, University of the Philippines Diliman

Directory of Open Access Journals (Sweden)

Sandra L. Yap

2013-12-01

Full Text Available The University of the Philippines Herbarium was established in 1908 and originally located in Ermita, Manila. The majority of its pre-war collections were destroyed during World War II, and no formal records of its specimens were preserved. Since then, multiple efforts to restore and improve the Herbarium have been proposed and implemented, most notably its move to the UP Diliman campus. In 1999, the Herbarium was off icially renamed as the Jose Vera Santos Memorial Herbarium after the noted grass expert, who initiated rehabilitation work in the Herbarium after the war. The Herbarium is registered with the international code PUH in the Index Herbariorum, a global directory of public herbaria managed by the New York Botanical Garden. To assess the accessioned (uniquely numbered and recorded collection of the Herbarium, an electronic database of its accessions was created.The Herbarium currently contains 14,648 accessions, 12,681 (86.6% of which were collected in the Philippines. This is comprised of 309 families, 1903 genera, and 4485 distinct species. Thirty-nine type specimens form part of the collection, only one of which is a holotype. On the basis of major plant groups, angiosperms make up 71% of the collection. Unsurprisingly, Family Poaceae has the largest number of specimens at 2,759 accessions. The earliest dated Philippine specimen was collected by E.D. Merrill in 1902, and roughly half of the total accessioned specimens were collected in the 1950s and 1970s. The two most prolif ic collectors were Santos and Leonardo L. Co, with 2,320 and 2,147 specimens, respectively. Luzon is the most well-represented island group with 2,752 specimens collected in Metro Manila alone. At present, PUH Curator James V. LaFrankie is working on the expansion of the collection and upgrading of the herbarium to encourage future educational and research activities.

Dosimetry of radium equivalent in construction material of brick works in Sao Jose do Sabugi City - Paraiba, Brazil

Energy Technology Data Exchange (ETDEWEB)

Araujo, Eduardo Eudes Nobrega de; Santos Junior, Jose Araujo dos; Amaral, Romilton dos Santos; Santos, Josineide Marques do Nascimento; Spacov, Isabel Cristina Guerra; Fernandez, Zahily Herrero, E-mail: eduardo.eudes@ufpe.br, E-mail: jaraujo@ufpe.br, E-mail: romilton@ufpe.br, E-mail: neideden@hotmail.com, E-mail: isabelspacov@gmail.com, E-mail: zahily1985@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear. Grupo de Radioecologia

2015-07-01

The earth's crust has in its composition the Naturally Occurring Radioactive Material (NORM) that may have increased concentration due to activities of exploration and extraction of environmental resources. The civil construction is an economic activity that requires the use of much of the natural resources, such as the raw material of brick works, like clays, mainly used for the production of bricks and tiles. These construction materials may contain high levels of natural radioactive elements, even with concentrations higher than the limits established, given that the levels vary according to the composition of rocks and soil, due to the geological formation and may result in increased exposure of humans to natural radioactive activities. In this context, the radioecological dosimetry is defined in terms of Radium Equivalent activity (Ra{sub eq}), that ensure radiometric conditions for the use of material derived from clays before its final application in housing construction, an initiative that ensures the radioecological safety of population. Thus, this study aimed to establish the calculation of Ra{sub eq} in the raw material of brick works located in Sao Jose do Sabugi city, state of Paraiba, in an area adjacent to the uranium deposits of Espinharas, to estimate the risks associated with primordial radionuclides attributed to TENORM activities (Technologically Enhanced Naturally Occurring Radioactive Materials) from the extraction and use of clay as a raw material in the manufacture of bricks and tiles. Analyses were performed by High Resolution Gamma Spectrometry, with HPGe-Be detector, assuming the state of secular radioactive equilibrium. The results ranged from 183.2 to 747.78 Bq/kg, with an average of 494.6 Bq/kg which exceeded the limit of 370 Bq/kg established by UNSCEAR for construction materials. Some samples obtained values exceeded by up the double this limit, suggesting control and radiometric certification for application of this material

Spectral shift reactor control method

International Nuclear Information System (INIS)

Impink, A.J.

1982-01-01

A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

Reactor physics

International Nuclear Information System (INIS)

Ait Abderrahim, H.

1998-01-01

Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

Innovative hybrid biological reactors using membranes; Reactores biologico hibrido innovadores utilizando membranas

Energy Technology Data Exchange (ETDEWEB)

Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

2011-07-01

In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

Evaluation of the levels of particles PM10 and nitrogen dioxide at the city of San Jose, Costa Rica: 2005-2006

International Nuclear Information System (INIS)

Herrera-Murillo, Jorge; Rodriguez Roman, Susana

2009-01-01

The levels of particulate matter PM 10 were determined at two sites of San Jose City (Catedral Metropolitana and Junta de Educacion), during a year (September 2005-September 2006), obtaining as annual average 36 ± 8 μg/m 3 and 25 ± 7 μg/m 3 , respectively. Also, the levels of sulfates, nitrates and chlorides were measured for both sites. The annual averages have resulted not to be significantly different for both sites, with a level of significance of 5%. Three of the fourteen sites of measurement of the concentration of nitrogen dioxide in the city, have presented higher values to the recommendation of the World Health Organization, monitoring for one month. The principal component analysis that were applied to the data of this gas, has showed that the variations in the levels are due to large-scale phenomena (meteorological). However, the concentration of sulfate present in the particulate matter has reached higher values to those recorded in cities like Rio de Janeiro and Mexico D.F., product probably of the high sulfur content present in the fuels that are used by the vehicle fleet at Costa Rica. The principal component analysis indicates that has existed a strong correlation between the concentrations of sulfate and nitrate present in the particles PM 10 , pointing the contribution of anthropogenic sources. In the case of chloride, has highlighted the strong existing relationship with the meteorological parameters that were registered during the sampling period. (author) [es

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

International Nuclear Information System (INIS)

Joo, Young Sang; Lee, Jae Han

2007-02-01

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Lee, Jae Han

2007-02-15

The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

The Maple reactor project

International Nuclear Information System (INIS)

Malkoske, G.R.; Labrie, J.-P.

2003-01-01

MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Methanogenesis in Thermophilic Biogas Reactors

DEFF Research Database (Denmark)

Ahring, Birgitte Kiær

1995-01-01

Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process as ....... Experiments using biogas reactors fed with cow manure showed that the same biogas yield found at 550 C could be obtained at 610 C after a long adaptation period. However, propionate degradation was inhibited by increasing the temperature.......Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...

A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

Future view of total energy system and reactor engineering and reactor physics

International Nuclear Information System (INIS)

Ozawa, T.

1974-01-01

This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

International Nuclear Information System (INIS)

Bacher, P.

1997-01-01

A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

Research reactor support

International Nuclear Information System (INIS)

2005-01-01

Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

The prototype fast reactor

International Nuclear Information System (INIS)

Broomfield, A.M.

1985-01-01

The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

NCSU Reactor Sharing Program

International Nuclear Information System (INIS)

Perez, P.B.

1993-01-01

The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

International Nuclear Information System (INIS)

1975-06-01

The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

2012 review of French research reactors

International Nuclear Information System (INIS)

Estrade, Jerome

2013-01-01

Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

Upgradation of Apsara reactor

International Nuclear Information System (INIS)

Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

2009-01-01

Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

RA Reactor

International Nuclear Information System (INIS)

1989-01-01

This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

Climate-induced forest dieback as an emergent global phenomenon: Organized oral session at the Ecological Society of America/Society of Ecological Restoration Joint Meeting; San Jose, California, 5-10 August 2007

Science.gov (United States)

Allen, Craig D.; Breshears, David D.

2007-01-01

An organized oral session at the annual meeting of the Ecological Society of America in San Jose, Calif., posed this question: Is climate-induced drought stress triggering increasing rates and unusual patterns of forest die-off at a global scale? Twenty-nine researchers representing five continents reported on patterns, mechanisms, and projections of forest mortality.Observations include widespread forest dieback or reductions in tree cover and biodiversity in response to drought and warmer temperatures in the African Sahel (Patrick Gonzalez, The Nature Conservancy), Mediterranean and alpine Europe (Jorge Castro, Universidad de Granada), and Argentinean Patagonia (Thomas Kitzberger, Universidad Nacional del Comahue). In contrast, although much Eucalyptus mortality has resulted from recent droughts in Australia, warming trends have been less pronounced in the Southern Hemisphere and it is unclear if contemporary climate-induced tree mortality differs from previous historical drought impacts (Rod Fensham, Queensland Herbarium).

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

Materials for passively safe reactors

International Nuclear Information System (INIS)

Simnad, T.

1993-01-01

Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

The fast breeder reactor

International Nuclear Information System (INIS)

Davis, D.A.; Baker, M.A.W.; Hall, R.S.

1990-01-01

Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

Method of reactor operation

International Nuclear Information System (INIS)

Maeda, Katsuji.

1982-01-01

Purpose: To prevent stress corrosion cracks in stainless steels caused from hydrogen peroxide in reactor operation in which the density of hydrogen peroxide in the reactor water is controlled upon reactor start-up. Method: A heat exchanger equipped with a heat source for applying external heat is disposed into the recycling system for reactor coolants. Upon reactor start-up, the coolants are heated by the heat exchanger till arriving at a temperature at which the dissolving rate is faster than the forming rate of hydrogen peroxide in the coolants, and nuclear heating is started after reaching the above temperature. The temperature of the reactor water is increased in such a manner and, when it arrives at 140 0 C, extraction of control elements is started and the heat source for the heat exchanger is interrupted simultaneously. In this way spikes in the density of hydrogen peroxide are suppressed upon reactor start-up to thereby decrease the stress corrosion cracks in stainless steels. (Horiuchi, T.)

Reactor water sampling device

International Nuclear Information System (INIS)

Sakamaki, Kazuo.

1992-01-01

The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

Reactor power monitoring device

International Nuclear Information System (INIS)

Dogen, Ayumi; Ozawa, Michihiro.

1983-01-01

Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1975-11-01

Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

Reactor materials research as an effective instrument of nuclear reactor perfection

International Nuclear Information System (INIS)

Baryshnikov, M.

2006-01-01

The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

Matsuura, S.; Nakahara, Y.; Takano, H.

1983-09-01

Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Radiation shielding for fission reactors

Energy Technology Data Exchange (ETDEWEB)

Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

2000-03-01

Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

International Nuclear Information System (INIS)

Wittenbrock, N.G.

1982-01-01

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

Advanced reactor development: The LMR integral fast reactor program at Argonne

International Nuclear Information System (INIS)

Till, C.E.

1990-01-01

Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

Hybrid adsorptive membrane reactor

Science.gov (United States)

Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

2011-03-01

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

RB research reactor safety report

International Nuclear Information System (INIS)

Sotic, O.; Pesic, M.; Vranic, S.

1979-04-01

This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

Digital control of research reactors

International Nuclear Information System (INIS)

Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

1991-01-01

Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

LMFBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1988-01-01

Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

Fast breeder reactors

International Nuclear Information System (INIS)

Waltar, A.E.; Reynolds, A.B.

1981-01-01

This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of reactor theory, heat transfer, and fluid mechanics. The book is expected to be used most widely for a one-semester general course on fast breeder reactors, with the extent of material covered to vary according to the interest of the instructor. The book could also be used effectively for a two-quarter or a two-semester course. In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety. Methodology in fast reactor design and analysis, together with physical descriptions of systems, is emphasized in this text more than numerical results. Analytical and design results continue to change with the ongoing evolution of FBR design whereas many design methods have remained fundamentally unchanged for a considerable time

BWR type reactor

International Nuclear Information System (INIS)

Watanabe, Shoichi

1988-01-01

Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

Reactor core structure

International Nuclear Information System (INIS)

Higashinakagawa, Emiko; Sato, Kanemitsu.

1992-01-01

Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

Molten salt reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

2014-01-01

Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

Oklo natural reactor

International Nuclear Information System (INIS)

Fujii, Isao

1985-01-01

In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

Science.gov (United States)

Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

1994-01-01

This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

International Nuclear Information System (INIS)

Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

1991-01-01

The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

Reactor calculations for improving utilization of TRIGA reactor

International Nuclear Information System (INIS)

Ravnik, M.

1986-01-01

A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2002-01-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

Mirror fusion reactors

International Nuclear Information System (INIS)

Carlson, G.A.; Moir, R.W.

1978-01-01

We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

Mirror fusion reactors

International Nuclear Information System (INIS)

Anon.

1978-01-01

Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

Advanced CANDU reactors

International Nuclear Information System (INIS)

Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

1988-12-01

AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

International Nuclear Information System (INIS)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

Energy Technology Data Exchange (ETDEWEB)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

Conceptual design of reactor assembly of prototype fast breeder reactor

International Nuclear Information System (INIS)

Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

1996-01-01

The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

Fast breeder reactor research

International Nuclear Information System (INIS)

1975-01-01

Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

Reactor Coolant Pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Reactor coolant pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

United Kingdom and USSR reactor types

International Nuclear Information System (INIS)

Lewins, Jeffery

1988-01-01

The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

International Nuclear Information System (INIS)

Allen, Francis; Bonin, Hugues

2008-01-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

2008-07-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Reactor-core-reactivity control device

International Nuclear Information System (INIS)

Miura, Teruo; Sakuranaga, Tomonobu.

1983-01-01

Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

Reactor water level control device

International Nuclear Information System (INIS)

Utagawa, Kazuyuki.

1993-01-01

A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Breeding description for fast reactors and symbiotic reactor systems

International Nuclear Information System (INIS)

Hanan, N.A.

1979-01-01

A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

Nuclear reactor

International Nuclear Information System (INIS)

Miyashita, Akio.

1981-01-01

Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.

1976-01-01

A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

REACTOR SHIELD

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Reactor water quality degradation suppressing method upon reactor start up

International Nuclear Information System (INIS)

Maeda, Katsuharu.

1993-01-01

Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

Prometheus Project Reactor Module Final Report, For Naval Reactors Information

International Nuclear Information System (INIS)

MJ Wollman; MJ Zika

2006-01-01

The Naval Reactors Prime Contractor Team (NRPCT) led the development of a power plant for a civilian nuclear electric propulsion (NEP) system concept as part of the Prometheus Project. This report provides a summary of the facts, technical insights, and programmatic perspectives gained from this two-year program. The Prometheus Project experience has been extensively documented to better position the US for future space reactor development. Major Technological and engineering challenges exist to develop a system that provides useful electric power from a nuclear fission heat source operating in deep space. General issues include meeting mission requirements in a system that has a mass low enough to launch from earth while assuring public safety and remaining safely shutdown during credible launch accidents. These challenges may be overcome in the future if there is a space mission with a compelling need for nuclear power to drive development. Past experience and notional mission requirements indicate that any useful space reactor system will be unlike past space reactors and existing terrestrial reactors

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2001-07-15

Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

Reactor of the XXI century

International Nuclear Information System (INIS)

Zhotabaev, Zh.R.; Solov'ev, Yu.A.

2001-01-01

The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

Reactor core fuel management

International Nuclear Information System (INIS)

Silvennoinen, P.

1976-01-01

The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

International Nuclear Information System (INIS)

Duce, S.W.; Marley, A.W.; Freeman, A.L.

1989-12-01

Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity