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Sample records for alto lazio-2 reactor

  1. A model of Alto Lazio boiling water reactor using the LEGO code balance of plant simulation

    International Nuclear Information System (INIS)

    Spelta, S.; Garbossa, G.B.

    1989-01-01

    An extensive effort has been made at the Italian National Electricity Board (ENEL) to construct and validate a LEGO model capable of simulating the operational transients of the Alto Lazio Nuclear Station, a two twin-units site with BWR/6 class reactors, rated at 2894 MWt and with Mark III containment. LEGO is a modular package developed at the Research and development Department of the Italian National Electricity Board (CRA-ENEL) for computer aided modeling of fossil-fired and nuclear steam power plants. In this paper a system analysis model capable of describing steady-state and transient performance of the Balance of Plant (BOP) of the Alto Lazio Power Station is presented. This is one of two companion papers devoted to the description of the overall plant model including both the Nuclear Steam Supply System (NSSS) and the BOP. In the paper, after a brief summary of the main LEGO characteristics, a description of the BOP lay-out is presented. The overall model, which has been set-up, including control systems and automation, is very detailed and consists of almost 2000 differential or algebraic equations. After a brief description of the mathematical model, two significant transients obtained using the overall model are presented and discussed

  2. A model of Altio Lazio boiling water reactor using the LEGO code nuclear steam supply system simulation

    International Nuclear Information System (INIS)

    Garbossa, G.B.; Spelta, S.; Cori, R.; Mosca, R.; Cento, P.

    1989-01-01

    An extensive effort has been made at the Italian National Electricity Board (ENEL) to construct and validate a LEGO model capable of simulating the operational transients of the Alto Lazio Nuclear Station, a two-twin units site with BWR/6 class reactors, rated at 2894 MWt and with Mark III containment. The desired end-product of this effort is an overall plant model consisting of the Nuclear Steam Supply System model, described in this paper, and the Balance of Plant model, capable of simulating the transient response of Alto Lazio Station. The models utilize the in-house developed LEGO code, which is a modular package oriented to power plant modeling and suitable to perform transient analyses to assist during power plant design, control system design and operating procedure verification. The ability of the NSSS model to predict correctly the plant response is demonstrated through comparison with results calculated by the vendor, using REDY code, and by an in-house RETRAN-02 model

  3. Improved safety features in the design of Alto Lazio NPP

    International Nuclear Information System (INIS)

    Bava, G.; Cianciolo, T.; Del Nero, G.

    1988-01-01

    The ALTO LAZIO Nuclear Power Plant, two 1000Mwe units, is a BWR 6/MARK III located about 100 km north of Rome, on the Tyrrhenian Sea Coasts. The construction of the plant started in 1978, but it has recently been stopped by a Government decision following a national referendum, when the units were about 70% completed. This paper is mainly intended to illustrate the major safety features which have been implemented as result of specific requirements issued by the safety authority (ENEA DISP) during the construction permit stage or the subsequent licensing process. One of the tools used to identify the need for design modifications has been a comprehensive reliability analysis of safety system: in the paper the methods used and the major results obtained by this study are briefly presented. Also, the approach used in the investigation of severe accidents and major applications in the area of plant design and emergency procedures are briefly discussed; furthermore the trend toward a simpler mitigation concept is described

  4. Air and water pollution sources analysis in Northern Lazio; Censimento di fonti d'inquinamento atmosferico e idrico nell'Alto Lazio

    Energy Technology Data Exchange (ETDEWEB)

    Triolo, L; Barlattini, M; Sidoti, G; Tanzi, V; Testa, V [ENEA, Divisione Biotecnologie e Agricultura, Centro Ricerche Casaccia, Rome (Italy); Naviglio, L [ENEA, Divisione Protezione dell' Uomo e degli Ecosistemi, Centro Ricerche Casaccia, Rome (Italy); Boni, E; Proli, L; Santella, A; Squillacioti, T [Cooperativa Energia e Territorio Srl, Viterbo (Italy)

    2001-07-01

    Assessment of pollution sources was carried out in the North of Lazio because of ENEA interest to investigate on ecosystem and of this region. First part of the study concerns atmospheric emissions from oil combustion (SO{sub 2}, NO{sub x}, HC, CO, Particulate) associated to industrial and civil activities of Viterbo Province. Assessment of pesticide immission in environment from main crops of 9 omogeneous agricultural areas was completed. Pollution of Marta and Mignone rivers evaluation was carried out in the second part of the study. Agriculture chemical compounds and sewage chemicals wastes were estimated in Marta and Mignone basins. Inventory of industrial and civil emissions of atmospheric pollutants in Civitavecchia, S. Marinella and Tarquinia municipalities constitutes the third part of this study. Great relevance of atmospheric emissions has been attributed to thermoelectric plants of Torvaldaliga Sud, Torvaldaliga Nord and Fiumaretta and to cement factory of Italcementi, that are sited in Civitavecchia territory, Also a relevant contribution to atmospheric pollution is given by Civitavecchia port activities overall because of large oil products tanks. [Italian] L'analisi delle fonti inquinanti dei territori del Nord del Lazio e' stata effettuata per fornire una base dati per l'assessment degli effetti dell'inquinamento atmosferico sugli ecosistemi terrestri e acquatici e sugli agroecosistemi del territorio stesso. Nella prima parte dell'indagine sono state stimate, per l'intera provincia di Viterbo, le emissioni di particolato, SO{sub 2}, NO{sub x}, Idrocarburi e CO causate dalla combustione di prodotti petroliferi associati ai settori industriali e civili (in particolare al riscaldamento domestico e all'autotrasporto). Per l'agricoltura la valutazione delle emissioni di pesticidi e' stata effettuata sulla base delle pratiche agricole delle colture di olivo, vite, nocciolo e di altre orticole, fruttifere e cereali presenti in 9 aree omogenee rispetto all

  5. Air and water pollution sources analysis in Northern Lazio; Censimento di fonti d'inquinamento atmosferico e idrico nell'Alto Lazio

    Energy Technology Data Exchange (ETDEWEB)

    Triolo, L.; Barlattini, M.; Sidoti, G.; Tanzi, V.; Testa, V. [ENEA, Divisione Biotecnologie e Agricultura, Centro Ricerche Casaccia, Rome (Italy); Naviglio, L. [ENEA, Divisione Protezione dell' Uomo e degli Ecosistemi, Centro Ricerche Casaccia, Rome (Italy); Boni, E.; Proli, L.; Santella, A.; Squillacioti, T. [Cooperativa Energia e Territorio Srl, Viterbo (Italy)

    2001-07-01

    Assessment of pollution sources was carried out in the North of Lazio because of ENEA interest to investigate on ecosystem and of this region. First part of the study concerns atmospheric emissions from oil combustion (SO{sub 2}, NO{sub x}, HC, CO, Particulate) associated to industrial and civil activities of Viterbo Province. Assessment of pesticide immission in environment from main crops of 9 omogeneous agricultural areas was completed. Pollution of Marta and Mignone rivers evaluation was carried out in the second part of the study. Agriculture chemical compounds and sewage chemicals wastes were estimated in Marta and Mignone basins. Inventory of industrial and civil emissions of atmospheric pollutants in Civitavecchia, S. Marinella and Tarquinia municipalities constitutes the third part of this study. Great relevance of atmospheric emissions has been attributed to thermoelectric plants of Torvaldaliga Sud, Torvaldaliga Nord and Fiumaretta and to cement factory of Italcementi, that are sited in Civitavecchia territory, Also a relevant contribution to atmospheric pollution is given by Civitavecchia port activities overall because of large oil products tanks. [Italian] L'analisi delle fonti inquinanti dei territori del Nord del Lazio e' stata effettuata per fornire una base dati per l'assessment degli effetti dell'inquinamento atmosferico sugli ecosistemi terrestri e acquatici e sugli agroecosistemi del territorio stesso. Nella prima parte dell'indagine sono state stimate, per l'intera provincia di Viterbo, le emissioni di particolato, SO{sub 2}, NO{sub x}, Idrocarburi e CO causate dalla combustione di prodotti petroliferi associati ai settori industriali e civili (in particolare al riscaldamento domestico e all'autotrasporto). Per l'agricoltura la valutazione delle emissioni di pesticidi e' stata effettuata sulla base delle pratiche agricole delle colture di olivo, vite, nocciolo e di altre orticole, fruttifere e

  6. ENERGIA SOSTENIBILE: PIANIFICAZIONE STRATEGICA E PROGRAMMI ECONOMICI NELLA REGIONE LAZIO

    Directory of Open Access Journals (Sweden)

    Leonide Tocchi

    2017-06-01

    Full Text Available The new energy and regulatory scenarios on European and Italian level require a review of the regional energy strategies. Transitioning the global economy from fossil fuels to renewable energy sources has been identified as a key strategy for mitigating climate change. Energy sector transformation needs smart policies. The Lazio region is drawing up a new strategy for sustainable energy that aims to define the necessary conditions for development of a regional energy system increasingly turned to the use of renewable sources and efficient energy use as a means for greater environmental protection, in particular for the purpose of reduction of greenhouse gases (GHG. The strategy aims to facilitate the transition to a low carbon economy by increasing energy production from renewable sources, fostering a green economic recovery and the creation of green jobs in Lazio Region.

  7. Analysis of the organic horticultural market in Lazio; Analisi della filiera ortofrutticola biologica del Lazio

    Energy Technology Data Exchange (ETDEWEB)

    Letardi, A [ENEA, Divisione Biotecnologie e Agricoltura, Centro Ricerche Casaccia, Rome (Italy); Lumaca, P [Centro Ecologico di Dimostrazione Agraria, Rome (Italy); Grandi, C; Dominicis, L [Centro Ecologico di Dimostrazione Agraria/Associazione Italiana per l' Agricoltura Biologica, Lazio, Rome (Italy)

    2001-07-01

    In 1998 Agriculture and Biotechnology Division of ENEA (BIOAG), Ecological Centre for Extension Service (CEDA), and Italian Association for Biological Agriculture (AIAB) established a research collaboration on the limiting factors that regulate marketing of fresh biological products. Field research was carried out, starting at the end 1998 to 1999, on horticultural production, mainly by means a fellowship in agriculture factors that regulate marketing of fresh biological products. Results and conclusion of the study focuses critical steps regulating productions, transformation and distribution of biological agriculture and could be associated to general situation of this sector in Italy. Moreover attention should be put on the rapid evolution of this sector in the last months, with respect to research time duration, i.e., 1998-1999 years, because of food safety emergencies and legislative innovations issued by European Commission. [Italian] Nel 1998 una lunga collaborazione tra ricercatori della Divisione Biotecnologie ed Agricoltura dell'ENEA, del Ceda (Centro Ecologico di Dimostrazione Agraria) e dell'AIAB (Associazione Italiana per l'Agricoltura Biologica), grazie all'apporto finanziario di un imprenditore privato interessato allo sviluppo del settore, produsse un bando di concorso per una borsa di formazione e studio sperimentale per laureato in agraria con specializzato in materie economiche. Grazie a tale borsa e' stata realizzata, tra la fine del 1998 e il 1999, una indagine sulla filiera agroalimentare biologica del Lazio, finalizzata all'analisi dei punti critici che limitavano i segmenti della commercializzazione e della distribuzione del prodotto fresco. Nella discussione su principali problemi per lo sviluppo dell'agricoltura biologica in Italia, ed in particolare nel Lazio, tra i ricercatori delle strutture sopra menzionate era emersa infatti una carenza di dati sperimentali certi che potessero supportare una serie di considerazioni gia' da noi

  8. Risk of acquiring tick-borne infections in forestry workers from Lazio, Italy

    OpenAIRE

    2010-01-01

    Abstract The seroprevalence of antibodies to Borrelia burgdorferi and tick-borne encephalitis (TBE) virus was evaluated in a group of forestry rangers in the Lazio region of Italy. One hundred and forty-five forestry rangers and 282 blood donors were examined by two-tiered serological tests for B. burgdorferi and TBE virus. Information on occupation, residence, tick bites, outdoor leisure activities and other risk factors was obtained. The prevalence of IgG/IgM antibodies to B. bur...

  9. Analysis of the organic horticultural market in Lazio; Analisi della filiera ortofrutticola biologica del Lazio

    Energy Technology Data Exchange (ETDEWEB)

    Letardi, A. [ENEA, Divisione Biotecnologie e Agricoltura, Centro Ricerche Casaccia, Rome (Italy); Lumaca, P. [Centro Ecologico di Dimostrazione Agraria, Rome (Italy); Grandi, C.; Dominicis, L. [Centro Ecologico di Dimostrazione Agraria/Associazione Italiana per l' Agricoltura Biologica, Lazio, Rome (Italy)

    2001-07-01

    In 1998 Agriculture and Biotechnology Division of ENEA (BIOAG), Ecological Centre for Extension Service (CEDA), and Italian Association for Biological Agriculture (AIAB) established a research collaboration on the limiting factors that regulate marketing of fresh biological products. Field research was carried out, starting at the end 1998 to 1999, on horticultural production, mainly by means a fellowship in agriculture factors that regulate marketing of fresh biological products. Results and conclusion of the study focuses critical steps regulating productions, transformation and distribution of biological agriculture and could be associated to general situation of this sector in Italy. Moreover attention should be put on the rapid evolution of this sector in the last months, with respect to research time duration, i.e., 1998-1999 years, because of food safety emergencies and legislative innovations issued by European Commission. [Italian] Nel 1998 una lunga collaborazione tra ricercatori della Divisione Biotecnologie ed Agricoltura dell'ENEA, del Ceda (Centro Ecologico di Dimostrazione Agraria) e dell'AIAB (Associazione Italiana per l'Agricoltura Biologica), grazie all'apporto finanziario di un imprenditore privato interessato allo sviluppo del settore, produsse un bando di concorso per una borsa di formazione e studio sperimentale per laureato in agraria con specializzato in materie economiche. Grazie a tale borsa e' stata realizzata, tra la fine del 1998 e il 1999, una indagine sulla filiera agroalimentare biologica del Lazio, finalizzata all'analisi dei punti critici che limitavano i segmenti della commercializzazione e della distribuzione del prodotto fresco. Nella discussione su principali problemi per lo sviluppo dell'agricoltura biologica in Italia, ed in particolare nel Lazio, tra i ricercatori delle strutture sopra menzionate era emersa infatti una carenza di dati sperimentali certi che potessero supportare una serie di

  10. Research and development activity at ENEL for next generation reactors

    International Nuclear Information System (INIS)

    Fornaciari, P.

    1992-01-01

    Italy, a world leader in nuclear electricity production in the mid sixties, became the only large industrialized country which renounced nuclear energy; a fact hardly understandable for a country that, in addition has a worrisome dependence on imported foreign energy. The paper reports the political events that brought Italy to a condition not easy to be sustained in the long run, citing in particular: the so-called 'reconversion' of the Alto Lazio nuclear plant into a fossil fuelled plant, the closure of the Caorso plant after the encouraging outcome of an IAEA inspection, and finally, the contrasting resolutions voted on the matter by the two Parliament chambers. The paper then deals with the issues which need to be solved to allow a nuclear renaissance in Italy. The research program on future reactor designs has been carried out by ENEL (the Italian National Electricity Board) in the framework of a wide international co-operation with the ambitious goal to demonstrate that, even in the case of a severe accident, no population evacuation, nor any long term limitation as to ground utilization is to be planned. On this important issue a large international consensus is growing worldwide. The willingness of ENEL to remain part of the nuclear community has been also proved by its adhesion to the World Association of Nuclear Operators (WANO), created to enhance the operational safety of nuclear plants worldwide

  11. [Prevention in times of economic crisis and spending review. The Lazio Region as a study case].

    Science.gov (United States)

    Di Marco, Marco; Marzuillo, Carolina; De Vito, Corrado; Matarazzo, Azzurra; Massimi, Azzurra; Villari, Paolo

    2013-01-01

    With cutbacks being implemented across a wide range of social and government programs throughout Europe and the rest of the world, preventive services have become more vulnerable. In this context, it is essential to properly focus the debate on public healthcare expenditure, stressing that financing preventive services is not merely a cost, but an investment in citizen well-being as well as economic stability and development. In Italy indeed all seem to agree on three priorities: i) strengthening prevention activities; ii) reorganization of hospital care; and iii) reinforcement of primary care. A plenty of data are available in Italy from some recently published authoritative reports. Given that health policies should be driven by a solid evidence base, it is important to look at the available data to understand if these priorities are justified. The Lazio Region, which is particularly under pressure since it is one of the regions with a formal regional recovery plan (Piano di Rientro), was chosen as a case-study. In the Lazio Region public health care expenditure is particularly high, but the health care expenditure for prevention activities is among the lowest of the Italian Regions. Major weakness points documented by the essential levels of care indicators included recommended vaccinations coverage, oncological screening programs, residential beds for the elderly and persons with disability and hospital care efficiency. Avoidable mortality is higher in the Lazio than in the rest of the country, as well as the prevalence of some major behavioral risk factors. Even if all data available support the choice to consider prevention activities as a priority, it is essential to increasing the value of prevention, investing money in preventive interventions of proven effectiveness and cost-effectiveness and promoting synergies with institutions outside the health care sector, implementing in a more efficient way the principle of Health in All Policies.

  12. Twelve-year analysis of cattle and buffalo slaughtering in Lazio Region (2000-2012: animal husbandry and veterinary public health implications

    Directory of Open Access Journals (Sweden)

    Selene Marozzi

    2014-02-01

    Full Text Available In recent years, beef meat chain has undergone major transformations due to Community legislation and market changes. The purpose of this work is to analyse the information recorded in Banca Dati Nazionale (BDN; Italian computerised database for the identification and registration of bovine animals on cattle and buffaloes slaughtered between 2000 and 2012 and related to Lazio Region as a result of breeding and/or slaughtering place. The analysis of the data showed a negative trend (-20.7% for cattle slaughtered from 2000 to 2012. Most of this animals had been raised in Lazio Region (86% and in particular in the province of Frosinone. The average age at slaughter for female is about 4 years (1417 days and for males of 547 days. The buffaloes, however, are intended for slaughter at an average age of about 8 years, if female, and about one year if male.

  13. Trends of some wintering waterbirds in Lazio (1993-2006

    Directory of Open Access Journals (Sweden)

    Massimo Brunelli

    2012-09-01

    Full Text Available Since the 90s, censuses of wintering waterfowl have been carried out in the main wetlands of Lazio. We analysed the trends of 31 species in the 1993-2006 period (base year 1993 by means of TRIM (Trends and Indices Monitoring data software (Model 3. Among the species regularly recorded in the region, Ardea alba, Ardea cinerea, Bubulcus ibis and Anser anser showed a strong increase; Podiceps cristatus, Nycticorax nycticorax, Egretta garzetta, Phoenicopterus ruber, Anas penelope, Anas strepera, Anas crecca, Anas platyrhynchos, Anas clypeata, Netta rufina, Aythya ferina, Aythya nyroca, Circus aeruginosus, Fulica atra, Pluvialis apricaria and Vanellus vanellus showed a moderate increase; Gavia arctica, Tachybaptus ruficollis, Podiceps nigricollis, Phalacrocorax carbo, Aythya fuligula and Numenius arquata resulted “stable”; Botaurus stellaris, Tadorna tadorna, Anas acuta, Pluvialis squatarola and Calidris alpina showed an uncertain trend. The trends for most species are similar to those recorded at a national level.

  14. DUE NUOVE SPECIE DI OTIORHYNCHUS (LIXORRHYNCHUS REITTER, 1914 E UNA NUOVA SPECIE DI RAYMONDIONYMUS WOLLASTON, 1873 DEI MONTI AURUNCI (LAZIO (COLEOPTERA, CURCULIONOIDEA

    Directory of Open Access Journals (Sweden)

    Paolo Magrini

    2008-10-01

    Full Text Available Nella presente nota vengono descritti tre nuovi Curculionoidea ipogei dei Monti Aurunci (Lazio: Otiorhynchus (Lixorrhynchus avoni n. sp.; Otiorhynchus (Lixorrhyn­chus paulae n. sp. e Raymondionymus pulcherrimus n. sp. Nel testo vengono riportate immagini fotografiche dei principali caratteri esoscheletrici (sia interni che esterni che contraddistinguono le nuove specie, nell’ambito dei gruppi di appartenenza. Una cartina geografica riassume lo stato dell’attuale distribuzione dei Lixorrhynchus anoftalmi o microftalmi in Italia penisulare e nell’area Sardo-Corsa. Le prime due specie presentano indubbie affinità con Otiorhynchus (Lixorrhynchus bastianinii Magrini, Meoli & Abbazzi, 2005, recentemente descritto dei Monti Aurunci centrali [Grava dei Serini (= Grotta dei Serini 587 La/FR], mentre la terza specie costituisce, insieme a R. meggiolaroi (Osella, 1977 (Liguria, R. eximius Meregalli & Osella, 2006 (Lazio, Monti Simbruini e R. zoiai (Osella & Giusto, 1985 (Piemonte, Massiccio del Monviso, un gruppo immediatamente riconoscibile rispetto ai taxa congeneri, per la particolare conformazione del pronoto.

  15. Use of emergency department services by women victims of violence in Lazio region, Italy.

    Science.gov (United States)

    Farchi, Sara; Polo, Arianna; Asole, Simona; Ruggieri, Maria Pia; Di Lallo, Domenico

    2013-07-19

    Violence against women is a significant health problem and a hidden phenomenon, in Italy that about 31% of the women have been victims of violence once in life. Aims of this study are to describe characteristics of women victims of violence (VV) attending the EDs in the Lazio region in 2008 and to illustrate the frequency and characteristics of previous ED visits. Using the Emergency Information System, visits of women, (15-49 years), in the 60 EDs, for a violent trauma have been analysed. For each VV identified, we considered the last episode and searched for ED attendances in a six year period (2003-08) in order to identify other visits. We performed descriptive analyses of socio-demographic and clinical factors of VV and we analyzed the impact previous ED visits. We compared ED utilization of women VV with a random sample of women with the same age distribution who gave birth in 2008. In 2008, 7,725 ED attendances of women VV were found (1.1% of the ED visits) corresponding to 6,936 women (prevalence = 52.0x10,000). The mean number of ED visits for each woman in five years was 5.0 (1-190). Prevalent diagnoses were contusions (45.8%), neurotic disorders (5.4%) complications of medical care (6.3%). The women were young, approximately 70% were residents in Rome or the surrounding areas. Foreign women were three times more likely to visit the ED for intentional injuries than were Italian women (114.1 vs 44.4 per 10.000). This study shows high prevalence of violence against women in Lazio region, Italy. Most of the women have been visited by the ED several times before the violent episode, often with traumas. ED medical and nursing staff should be prepared and trained to successfully manage victims of violence.

  16. Rural Tourism and Local Development: Typical Productions of Lazio

    Directory of Open Access Journals (Sweden)

    Francesco Maria Olivieri

    2014-12-01

    Full Text Available The local development is based on the integration of the tourism sector with the whole economy. The rural tourism seems to be a good occasion to analyse the local development: consumption of "tourist products" located in specific local contexts. Starting from the food and wine supply chain and the localization of typical productions, the aim of the present work will be analyse the relationship with local development, rural tourism sustainability and accommodation system, referring to Lazio. Which are the findings to create tourism local system based on the relationship with touristic and food and wine supply chain? Italian tourism is based on accommodation system, so the whole consideration of the Italian cultural tourism: tourism made in Italy. The touristic added value to specific local context takes advantage from the synergy with food and wine supply chain: made in Italy of typical productions. Agritourism could be better accommodation typology to rural tourism and to exclusivity of consumption typical productions. The reciprocity among food and wine supply chain and tourism provides new insights on the key topics related to tourism development and to the organization of geographical space as well and considering its important contribution nowadays to the economic competitiveness.

  17. Sorveglianza della circolazione ambientale dei poliovirus nel Lazio

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    A.M. Patti

    2003-05-01

    Full Text Available Ancora oggi in tutto il mondo il vaccino antipolio più utilizzato è l’OPV costituito da virus viventi attenuati che vengono eliminati per un periodo di tempo variabile dal soggetto vaccinato. L’immissione di virus vaccinali nell’ambiente è stata in passato, e lo è tuttora nelle zone endemiche, estremamente importante per assicurare e la competizione con il poliovirus selvaggio e una immunità di gregge. Nei paesi polio-free, ed in futuro in tutto il mondo, la circolazione di virus vaccinali potrebbe viceversa diventare un punto critico in grado di inficiare i risultati dell’eradicazione. Infatti i virus vaccino derivati, replicando, retromutano verso la neurovirulenza e/o accumulano mutazioni che alla fine conferiscono loro caratteristiche del tutto diverse dai ceppi parentali; inoltre possono anche ricombinarsi con il selvaggio o con altri enterovirus assumendo caratteristiche di virulenza e di trasmissibilità interumana che emergono con lo scoppio di focolai epidemici. Obiettivo del presente progetto è stata la valutazione della circolazione dei poliovirus e degli eventuali virus vaccino derivati in matrici ambientali nella regione Lazio nel periodo 1996-2002. Metodo: sono stati analizzati 26 campioni di liquami e 36 campioni di acque superficiali contaminate da liquami. Le particelle virali sono state concentrate mediante ultra filtrazione tangenziale (10.000 NMWR – Millipore. I concentrati sono stati seminati su cellule BGM ed L20B. I virus isolati sono stati identificati con antisieri specifici (RIUM e sui poliovirus, presso l’ISS, sono stati effettuati la differenziazione intratipica, il sequenziamento della regione VPI/2A, il sequenziamento della regione 5’ NCR e la regione codificante la polimerasi virale. Risultati e conclusioni: sono stati isolati complessivamente 6 poliovirus di cui 4 da acque superficiali. I virus erano tutti Sabin-kike e retromutati ma non ricombinanti. I dati ottenuti sottolineano l

  18. Reactor BR2

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2000-07-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

  19. Reactor BR2

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  20. Criterios para identificar patolog?as de alto costo en Colombia

    OpenAIRE

    Cuenta de alto costo, MinSalud

    2010-01-01

    Al identificar posibles pacientes de alto costo se debe definir si existen caracter?sticas que determinan su comportamiento como pacientes de alto costo, para definir si dicha patolog?a puede considerarse como Enfermedad de Alto Costo en Colombia.

  1. [Epidemiology of work-related accidents in the Lazio Region of Italy].

    Science.gov (United States)

    Marchetti, Aurora; Mantovani, Jessica; Di Lallo, D; Di Napoli, A; Guasticchi, Gabriella

    2011-01-01

    Prevention of work-related accidents requires an in-depth epidemiological assessment of the issue. In Italy the most used databases are from the national insurance (INAIL) and research (ISPESL) institutes. However, these data are only available several years after the time of accident. To describe the characteristics of accidents and evaluate factors potentially associated with hospitalization using the Information System of Hospital Emergency Departments (SIES). We analyzed 51.705 Emergency Department (ED) work-related accident admissions in the Lazio Region of Italy in 2008 among workers aged 16-65 years. Information on socio-demographics, diagnosis, triage codes, and outcome of ED admissions were gathered. We performed a logistic regression model to estimate association between these factors and risk of hospitalization after ED admission. The subjects' mean age was 39.1 (SD 11.0); 71.5% woere men, 12.7% were foreigners, 5.9% arrived by ambulance, 4.5% with triage red/yellow tags, 2.7% were hospitalized. Diagnosis was trauma in 85.1%, orthopaedic lesions in 8.3%. We found a higher risk of hospitalization in subjects with: one year of age increase (OR=1.02; 95% CIs: 1.01-1.03), males (OR=1.68; 95% CIs: 1.44-1.97), foreigners coming from countries with high emigration rates (OR=1.55; 95% CIs: 1.31-1.82), ED triage red/yellow tags (OR=84.47; 95% CIs: 47.06-151.60). It was confirmed that data fr-om an emergency health care information system can be a useful complement to information gathered by national insurance and research institutes, thus resolving the limit posed by the delay in availability for analysis of these data after the occurrence of accidents. We also identified some factors potentially associated with more serious accidents, which constitute a basis for planning and implementing specific public health preventive interventions.

  2. The areal reduction factor: A new analytical expression for the Lazio Region in central Italy

    Science.gov (United States)

    Mineo, C.; Ridolfi, E.; Napolitano, F.; Russo, F.

    2018-05-01

    For the study and modeling of hydrological phenomena, both in urban and rural areas, a proper estimation of the areal reduction factor (ARF) is crucial. In this paper, we estimated the ARF from observed rainfall data as the ratio between the average rainfall occurring in a specific area and the point rainfall. Then, we compared the obtained ARF values with some of the most widespread empirical approaches in literature which are used when rainfall observations are not available. Results highlight that the literature formulations can lead to a substantial over- or underestimation of the ARF estimated from observed data. These findings can have severe consequences, especially in the design of hydraulic structures where empirical formulations are extensively applied. The aim of this paper is to present a new analytical relationship with an explicit dependence on the rainfall duration and area that can better represent the ARF-area trend over the area case of study. The analytical curve presented here can find an important application to estimate the ARF values for design purposes. The test study area is the Lazio Region (central Italy).

  3. L' influenza nella regione Lazio dal 1999 al 2003: casi di sindrome influenzale, ricoveri ospedalieri per malattie respiratorie e coperture vaccinali

    Directory of Open Access Journals (Sweden)

    A. Pasquarella

    2003-05-01

    Full Text Available

    Obiettivi: 1 Descrivere l’andamento dei ricoveri ospedalieri per patologie respiratorie acute e croniche concomitanti alle epidemie stagionali da virus influenzale dal 1999 al 2003, in relazione con la segnalazione dei casi di sindrome influenzale (ILI da parte dei medici sentinella. 2 Misurare l’eccesso dell’ospedalizzazione influenza-correlata nelle diverse fasce di età rispetto ai periodi non epidemici. 3 Analizzare le modificazioni del ricorso al ricovero ospedaliero in relazione al tasso di copertura della vaccinazione antinfluenzale nella popolazione anziana, su scala regionale e nelle diverse ASL.

    Metodi: sono stati estratti dal Sistema Informativo
    Ospedaliero i ricoveri per patologie respiratorie
    influenza-correlate (codici ICD9-CM: 480-487; 460-
    466; 490-496 relativi agli anni 1999-2003.
    L’incidenza di ILI è stata stimata sulla base delle
    segnalazioni dei medici sentinella afferenti alla
    rete FLU-ISS dell’Istituto Superiore di Sanità.

    Per il calcolo dei tassi di copertura è stato utilizzato l’archivio
    nominativo dei soggetti vaccinati contro l’influenza,
    attivo nella regione Lazio dal 1999. Nel periodo considerato sono stati messi in relazione i tassi di ospedalizzazione età-specifici, le incidenze di ILI e le coperture vaccinali. L’eccesso di ospedalizzazione è stato misurato confrontando i tassi relativi ai periodi epidemici e non epidemici.

    Risultati: i tassi di ospedalizzazione per malattie respiratorie sono risultati costantemente superiori nei periodi di maggiore circolazione virale, in particolare negli ultrasessantaquattrenni. Con il progressivo aumento del tasso di copertura vaccinale regionale (da circa il 25% della stagione 1999-2000 a oltre il 60% della stagione 2002-2003 non si è registrata una corrispondente diminuzione dei ricoveri ospedalieri per patologie influenza-correlate.
    L

  4. Delay in diagnosis of pulmonary tuberculosis: a survey in the Lazio region, Italy

    Directory of Open Access Journals (Sweden)

    Patrizio Pezzotti

    2014-11-01

    Full Text Available OBJECTIVE: To estimate patient and health care delays in the diagnosis of PTB and to evaluate associated factors.METHODS: PTB incident cases ≥18 years diagnosed between September 2010 and September 2011 in the Lazio region; information on symptoms and date of onset, health professionals contacts, diagnostic exams performed, and drugs prescribed before diagnosis were collected through a standardized questionnaire. The total delay (TD was divided into patient delay (PD: from symptoms onset to first contact with healthcare services and health system delay (HSD: from first contact to diagnosis.RESULTS: 278 cases were evaluated. Median PD,HSD, and TD, were 31, 15, and 77.5 days, respectively. The median PD, HSD, and TD were significantly lower in foreign born patients (26, 10.5, 63.5, vs. 45, 36, 100 days, respectively. Other factors independently associated with longer delay were: absence of fever and presence of weight loss for PD; prior unspecific treatment, absence of cough, consult with a general practitioner, visit to an outpatient clinic, and a PD <30 days for HSD.CONCLUSIONS: In Italy, the delay in TB diagnosis is similar to that estimated in other European countries. Results indicate that actions aimed to reduce diagnostic delay should be primarily addressed to Italian patients.

  5. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  6. Reactor BR2: Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2000-01-01

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

  7. 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

  8. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  9. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2001-01-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  10. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  11. Safety features of TR-2 reactor

    International Nuclear Information System (INIS)

    Tuerker, T.

    2001-01-01

    TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

  12. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2002-01-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

  13. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  14. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    Moons, F.

    2007-01-01

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  15. Structural response of a nuclear power plant steel containment under H2 detonation

    International Nuclear Information System (INIS)

    Maresca, G.; Milella, P.P.; Pino, G.

    1993-01-01

    To get a better understanding of the containment wall behaviour under a detonation a simple but complete model is analysed in order to study the fluid-structure interaction during the explosion. The structure is represented by a single degree of freedom (SDOF) elastic-plastic system. This system is coupled to a monodimensional model of the containment atmosphere excited by hydrogen bursting. The atmosphere modeling allows to represent the shock propagation and the reflected wave effects. In the model a cylindrical geometry is used as reference. The obtained results are compared with data adopted in Italy to assess the structural integrity of the Alto Lazio NPP steel containment in the case of a severe accident. The limits of the model as well as the possible extensions are discussed in the paper together with a possible application in an experimental program directed to the assessment of failure criteria under severe accident conditions. (orig./HP)

  16. Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, Edgar

    2011-01-01

    The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

  17. 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

  18. 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

  19. Diseño de una Fuente de Alto Voltaje

    Directory of Open Access Journals (Sweden)

    José Enrique Eirez Izquierdo

    2013-10-01

    Full Text Available Este documento presenta las experiencias en el diseño de una fuente de alto voltaje, basada en multiplicadores de media onda. La fuente garantizará un voltaje de salida en el orden de 102 V y una corriente en el orden de 10-3 A. Se muestran y analizan resultados experimentales encaminados a su aplicación en la alimentación de un generador de pulsos de alto voltaje.

  20. Core monitoring at the WNP-2 reactor

    International Nuclear Information System (INIS)

    Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

    1992-01-01

    The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

  1. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

  2. Evaluation of the health effects of the new driving penalty point system in the Lazio Region, Italy, 2001–4

    Science.gov (United States)

    Farchi, Sara; Chini, Francesco; Rossi, Paolo Giorgi; Camilloni, Laura; Borgia, Piero; Guasticchi, Gabriella

    2007-01-01

    Objective The penalty point system was introduced in Italy in June 2003. The aim of this study was to evaluate the health effects of this legislation in the Lazio region. Methods Poisson models were used to compare emergency department visits, hospitalizations and death between the pre‐law and post‐law periods (July 2001–June 2003; July 2003–June 2004). Results The emergency department visit rate ratio (RR) of the two periods was 0.87 (95% confidence interval (CI) 0.86 to 0.88); the corresponding hospital admission RR was 0.87 (95% CI 0.84 to 0.9). The death RR was 0.93 (95% CI 0.82 to 1.05). Conclusion After the legislation was introduced, there were fewer visits to the emergency department, hospitalizations and death from road traffic injuries. However, the effect was lower than expected, and it decreased over time. PMID:17296692

  3. Avaliação do estado nutricional e da composição corporal das crianças índias do Alto Xingu e da etnia Ikpeng Nutritional status and body composition of two South American native populations - Alto Xingu and Ikpeng

    Directory of Open Access Journals (Sweden)

    Ulysses Fagundes

    2004-12-01

    Full Text Available OBJETIVOS: Avaliar o estado nutricional e a composição corporal de crianças índias das populações alto-xinguana e Ikpeng, comparando as populações. MÉTODOS: Avaliamos 95 crianças do Alto Xingu e 69 Ikpeng com idades entre 24 e 117 meses. Obtivemos dados sobre idade, peso, estatura, pregas cutâneas, circunferência do braço e impedância bioelétrica. Calculamos escores z para peso, estatura e estimativas da composição corporal. Tendo como referência o NCHS 2000, determinamos diagnóstico de baixo peso e baixa estatura como sendo inferior a -2 escores z para os indicadores peso/idade ou índice de massa corporal/idade e estatura/idade, respectivamente. Para obesidade, o ponto de corte foi 2 escores do indicador índice de massa corporal/idade. As massas corporais magra e gordurosa foram calculadas a partir de duas equações validadas na literatura. RESULTADOS: Diagnosticamos baixa estatura em 8,4% das crianças do Alto Xingu e em 37,7% das Ikpeng (p OBJECTIVES: To assess the nutritional and body composition of two Brazilian indigenous populations by comparing their nutritional status. METHODS: 95 children from Alto Xingu and 69 from Ikpeng were evaluated, ages ranged from 24 to 117 months. The study was performed in the Xingu Indigenous Park. Data collected were: age, weight, height, skin folds, arm circumference, resistance and reactance. The z-scores were calculated and classified according to the parameters defined by the National Center for Health Statistics (NCHS 2000. Shortness was defined as length or stature below -2, underweight as body mass index below -2, and overweight as body mass index above 2. RESULTS: Among children from Alto Xingu, the prevalence of shortness was 8.4%, while among Ikpengs the prevalence was 37.7% (p < 0.001. Underweight was diagnosed in 12.5% of Ikpeng's children. Values of fat-free mass were greater for children from Alto Xingu and no case of obesity was found. CONCLUSION: In this study, Ikpeng

  4. ¿A qué atribuyen el alto rendimiento escolar los estudiantes de buen rendimiento escolar proveniente de liceos con altos indices de vulnerabilidad?

    OpenAIRE

    Morales, Mario; Sepúlveda, Martitza

    2016-01-01

    La presente investigación tiene como finalidad comprender desde las subjetividades de los participantes egresados de secundaria, provenientes de instituciones escolares con altos índices de vulnerabilidad, los principales factores que han contribuido en la obtención de su alto rendimiento escolar. Son varios los modelos que se han utilizados para explicar el abandono de los estudiantes en los primeros años de universidad (Ethington, 1990; St. John, Cabrera y Asker, 2000; Spady, 1970; Braxton,...

  5. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  6. Casas de altos en Concepción

    Directory of Open Access Journals (Sweden)

    Rodrigo Fischer Pérez

    1990-06-01

    Full Text Available La casa de altos es una construcción eminentemente urbana. Pareada junto a otras construcciones, va formando bordes continuos, cuyos primeros pisos son generalmente comerciales y los superiores habitacionales.

  7. Liderazgo servidor y equipos de alto desempeño

    OpenAIRE

    Mejía Villegas, Estefanía

    2015-01-01

    El objetivo de este ensayo es mostrar cómo el liderazgo servidor influye positivamente en el desarrollo de los equipos de alto desempeño. En el ensayo se desarrollan la temática del liderazgo servidor y su principal objetivo, a la vez que se describen las características de un equipo de alto desempeño y sus etapas de desarrollo, cómo son influenciadas por el liderazgo servidor y cómo este modelo puede cumplir con sus requerimientos. 

  8. BR2 Reactor: Irradiation of fuels

    International Nuclear Information System (INIS)

    Verwimp, A.

    2005-01-01

    Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

  9. Ageing management of the BR2 research reactor

    International Nuclear Information System (INIS)

    Verpoortem, J. R.; Van Dyck, S.

    2014-01-01

    At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

  10. Altos penachos de escarcha

    OpenAIRE

    Josa, Lola; Lambea, Mariano

    2008-01-01

    El documento contiene la composición titulada “Altos penachos de escarcha”, perteneciente al Manojuelo Poético-Musical de Nueva York, recopilación manuscrita de piezas poético-musicales de los siglos XVII y XVIII que se conserva en la biblioteca de The Hispanic Society of America (New York) bajo la signatura Ms. HC. 380/821a. Se ofrece la partitura con la transcripción musical a notación moderna, la edición anotada del poema y todos aquellos datos que ha sido posible averiguar sobre cada piez...

  11. 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

  12. De-hospitalization of the pediatric day surgery by means of a freestanding surgery center: pilot study in the lazio region

    Directory of Open Access Journals (Sweden)

    Mangia Giovanni

    2012-02-01

    Full Text Available Abstract Background Day surgery should take place in appropriate organizational settings. In the presence of high volumes, the organizational models of the Lazio Region are represented by either Day Surgery Units within continuous-cycle hospitals or day-cycle Day Surgery Centers. This pilot study presents the regional volumes provided in 2010 and the additional volumes that could be provided based on the best performance criterion with a view to suggesting the setting up of a regional Freestanding Center of Pediatric Day Surgery. Methods This is an observational retrospective study. The activity volumes have been assessed by means of a DRG (Diagnosis Related Group-specific indicator that measures the ratio of outpatients to the total number of treated patients (freestanding indicator, FI. The included DRGs had an FI exceeding the 3rd quartile present in at least a health-care facility and a volume exceeding 0.5% of the total patients of the pediatric surgery and urology facilities of the Lazio Region. The relevant data have been provided by the Public Health Agency and relate to 2010. The best performance FI has been used to calculate the theoretical volume of transferability of the remaining facilities into freestanding surgery centers. Patients under six months of age and DRGs common to other disciplines have been excluded. The Chi Square test has been used to compare the FI of the health-care facilities and the FI of the places of origin of the patients. Results The DRG provided in 2010 amounted to a total of 5768 belonging to 121 types of procedures. The application of the criteria of inclusion have led to the selection of seven final DRG categories of minor surgery amounting to 3522 cases. Out of this total number, there were 2828 outpatients and 694 inpatients. The recourse of the best performance determines a potential transfer of 497 cases. The total outpatient volume is 57%. The Chi Square test has pointed to a statistically significant

  13. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    Neve de Mevergnies, M.

    1977-01-01

    The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

  14. Fission product release from SLOWPOKE-2 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

    1994-12-31

    Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

  15. Production and study of fission fragments, from Lohengrin to Alto; Production et etude des fragments de fission, de Lohengrin a Alto

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, F

    2005-06-15

    The study of nuclei far from stability is constitutive of the history of nuclear physics at its very beginning and has been making considerable great strides since then. The study of these nuclei give the opportunity to reach new information on the nuclear structure and thus to measure the solidity of our knowledge on nuclear matter and its validity when it is pushed to its limits. The reaction selected for the production of exotic nuclei in the framework of the PARRNe program is the fission of uranium 238. The nuclei produced have an intermediate mass and are very rich in neutrons. The technique to recover them in order to accelerate them is the thick target method called also the Isol technique. The installation of the ancient Lep injector at the Tandem line in Orsay (IPN) is expected to increase by a factor 100 the production rate of exotic nuclei in the PARRNe program, it is the Alto project. The work presented here concerns studies carried out at the Lohengrin spectrometer installed at the ILL in Grenoble, and at the Tandem installation in Orsay. This document is divided into 4 parts: 1) in flight techniques at Lohengrin, 2) the Isol technique, 3) magic numbers in the domain N=50, and 4) the Alto project.

  16. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  17. Esporte de alto rendimento: reflexões psicanalíticas e utópicas

    Directory of Open Access Journals (Sweden)

    Mariana Hollweg Dias

    2012-01-01

    Full Text Available Este artigo busca fazer uma análise a respeito do esporte de alto rendimento a partir dos referenciais teóricos da Psicanálise e dos Estudos Utópicos, partindo do princípio de que a lógica do esporte de alto rendimento na contemporaneidade reverbera a lógica do laço social. A exigência da "alta performance" sempre é uma das características de nossa época que estão fortemente presentes no discurso do esporte de alto rendimento e que muitas vezes são fonte de padecimento para os sujeitos, atletas ou não. Apesar disso, o esporte ainda tem muito a contribuir na nossa sociedade, e a aposta deste trabalho é no que foi chamado utopia esportiva, que preconiza o acento na busca da superação mais do que o resultado final necessariamente no lugar mais alto do pódio.

  18. [Outsourcing of nursing human resources from ethical to management: a survey in a Lazio Region Hospital].

    Science.gov (United States)

    d'Amore, Maurizia; Peroni, Antonia

    2008-01-01

    In 2006, 279 nurses of a Lazio Region Hospital were assessed to verify whether certain Human Resources decisions, such as outsourcing, can negatively influence their working motivation and sense of belonging to a health organization and whether any dissatisfaction can be attributed to poor ethical information within the health service. The research method had a descriptive basis and for data collection a questionnaire with 35 questions was issued. Results showed that nurses felt strongly involved in the study and interesting aspects for management of human resources emerged, depicting an organization lacking in motivation : this confirmed one of the aspects of the study : poor levels of motivation and sense of belonging can be correlated to insufficient ethical information in local health organizations. The main working needs that emerged among the nurses of this hospital regarded economical retribution (90%), security and success (88%) , belonging (86%) and self-satisfaction (77%): the need for power was relatively low (40%). The strong points of the study were : the strong involvement of nurses, the value of the information gathered regarding working motivation, sense of belong to a nursing organization , working needs , ethical information in health environments, organization according to targets. The limits of the study were: limited number of nurses in outsourcing at the time of the study (12%), impossibility of comparing the results with data prior to the outsourcing choices made by the hospital in question.

  19. Advances in Reactor Physics, Mathematics and Computation. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

  20. 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

  1. Tracking of the LAZIO region shoreline from orthophotos AGEA 2014 and implementation of the database layer

    Science.gov (United States)

    Biscotti, Erik; Pizzeghello, Nicola; Murri, Chiara; Colistra, Graziano; Batzu, Ilenia

    2018-05-01

    The integrated coastal zone management (ICZM) is the modern approach used in the study, management and exploitation of the coastal area in various applications whereas in this area are concentrated interests concerning the most different fields, economic, environmental, legal, scientific and social. The coast is in fact inherently unstable by nature and consequently its characterization should take into account a continuous monitoring and updating of its variations and trends. The coastal area is that portion of land emerged and submerged containing the shoreline and is subject to both continental and marine geomorphic processes. The shoreline is the clearest expression of how this sector is particularly dynamic. Proper analysis and representation of the shape and nature of the coastal area are a first step to provide reliable and comparable tools to those who study and manage it. This paper presents the results of a study aimed to the realization of an integrated approach in the extraction of the shoreline using a case study of Lazio coast as a part of the European Project "Intercoast". This work is based on national and international directives on the coastal zone, whether linked to a more terrestrial or maritime area, still within the broad definition of Hydrography provided by the International Hydrographic Organization (IHO). The spatial information extracted by direct or indirect measurements of the most dynamic coastal sector emerged and submerged (emerged coast and sea bottom) have been provided by associating with a budget of measurement uncertainties, and assessing the quality.

  2. TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

    International Nuclear Information System (INIS)

    Kaczynski, G.M.; Woodruff, R.W.

    1985-12-01

    TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

  3. Once-through CANDU reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

  4. Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

    International Nuclear Information System (INIS)

    Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

    1988-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

  5. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  6. Tromboembolismo pulmonar masivo de alto riesgo asociado a foramen oval permeable

    Directory of Open Access Journals (Sweden)

    Antonio Miranda

    2012-04-01

    Full Text Available La alta mortalidad de los pacientes con tromboembolismo pulmonar masivo de alto riesgo amerita un enfoque terapéutico enérgico e invasivo que incluya la embolectomía pulmonar quirúrgica en aquellos pacientes con contraindicación para trombolisis o trombolisis fallida. Describimos un caso de tromboembolismo pulmonar masivo de alto riesgo que recibió tratamiento quirúrgico en vez de trombolisis debido a que al momento del diagnóstico presentaba un trombo móvil a través de un foramen oval permeable con altísima posibilidad de embolismo paradójico arterial.

  7. y El Alto, Bolivia

    Directory of Open Access Journals (Sweden)

    León Darío Parra Bernal

    2013-01-01

    Full Text Available En el presente caso de estudio se analiza la empresarialidad informal como un reto de política pública y económica. Para ello, se efectuaron 20 entrevistas en profundidad a microempresarios y comerciantes del sector informal en las ciudades de La Paz y El Alto, en Bolivia en 2010, y a 3 funcionarios públicos de instituciones de apoyo al fomento empresarial en el mismo país. La principal reflexión giró en torno al establecimiento de que los empresarios informales poseen un elevado nivel de influencia en la efectividad de las políticas públicas implementadas para su sector, así como en los mecanismos que se han utilizado en Bolivia para incluirlos en el proceso.

  8. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

  9. IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

  10. Calidad de los servicios de anticoncepción en El Alto, Bolivia The quality of contraception services in El Alto, Bolivia

    Directory of Open Access Journals (Sweden)

    Carmen Velasco

    1999-06-01

    Full Text Available El presente estudio tuvo por objetivo evaluar la calidad de los servicios de anticoncepción en la ciudad de El Alto, Bolivia. En su diseño se han contemplado cuatro elementos: 1 las relaciones entre los proveedores de servicios y sus clientes, 2 la disponibilidad de métodos anticonceptivos, 3 las condiciones de los servicios, y 4 la satisfacción de las usuarias. También se han tenido en cuenta las opiniones de los proveedores y de las usuarias y no usuarias de estos servicios, quienes se clasificaron como gubernamentales o no gubernamentales, de acuerdo con la administración de la institución a la que pertenecían. Los datos provinieron de un análisis de la situación de dichos servicios y de testimonios obtenidos de las participantes durante 1995. En cuanto a las relaciones interpersonales, se encontró que los proveedores percibían el trato del médico más favorablemente que las clientas, en tanto que las no usuarias lo percibían desfavorablemente. La percepción de un trato igualitario se correlacionó positivamente con la vestimenta que usaban las clientas. En cuanto a la disponibilidad de los métodos anticonceptivos, 15 de las 36 instituciones encuestadas no disponían de métodos modernos, a pesar de la existencia de una política nacional para proveerlos a la población. La oferta de estos servicios a parejas y a adolescentes es escasa, principalmente en las instituciones gubernamentales. El análisis de las condiciones de los servicios demostró que en algunas instituciones había problemas graves en la provisión de una atención de mínima calidad. Finalmente, este trabajo describe cómo la mayoría de estas limitaciones en la prestación de servicios de anticoncepción en El Alto pueden subsanarse mediante estrategias de costo moderado.The objective of this study was to evaluate the quality of contraception services in the city of El Alto, Bolivia. In the study design, four components were considered: 1 interpersonal

  11. Dislipidemias en comunidades pehuenches de Alto Biobio chileno Dyslipidemias in Pehuenche communities from Chilean Alto Bio Bio

    Directory of Open Access Journals (Sweden)

    Claudia Navarrete Briones

    2013-01-01

    Full Text Available Se realizó un estudio descriptivo y transversal de 400 habitantes (mayores de 15 años de edad de las comunidades pehuenches de Alto Biobio en Chile, de mayo a octubre del 2011, a fin de determinar la prevalencia de dislipidemias en esta población. La información necesaria se recolectó sobre la base de la normativa y los criterios del Ministerio de Salud y como resultados generales de las concentraciones plasmáticas promedio y la prevalencia de dislipidemias figuraron: colesterol total de 169,20 ±26,36 mg/dL y 8,2 %; lipoproteínas de baja densidad de 89,93 ±23,31 mg/dL y 4,5 %; triglicéridos de 145,89 ±48,96 mg/dL y 53,0 %; y lipoproteínas de alta densidad de 50 ±8,87 mg/dL y 28,3 %. Las cifras fueron inferiores en el grupo etario de 15-24 años y en personas de ascendencia pehuenche, con una pobre asociación a sobrepeso u obesidad abdominal; en general, resultaron menores a las de los citadinos.A descriptive and cross-sectional study was carried out in 400 people (over 15 years from Pehuenche communities of the Chilean Alto Biobio, from May to October 2011, in order to determine the prevalence of dyslipidemias in this population. Necessary information was collected on the basis of regulations and criteria of the Ministry of Health, and as general results of average plasma levels and prevalence of dyslipìdemia were: total cholesterol 169.20 ± 26.36 mg/dL and 8.2%; low-density lipoproteins 89.93 ± 23.31 mg/dL and 4.5%; triglycerides 145.89 ± 48.96 mg/dL and 53.0%; and high-density lipoproteins 50 ±8.87 mg/dL and 28.3%. The values were lower in the age group of 15-24 years and in Pehuenche people with poor association with abdominal obesity or overweight; in general, they were lower than those of the city people.

  12. Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

    International Nuclear Information System (INIS)

    Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

    1981-05-01

    The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

  13. Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

    1998-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

  14. Active species in a large volume N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

    2007-01-01

    A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

  15. G 2 reactor project; Projet de pile a double fin: G 2

    Energy Technology Data Exchange (ETDEWEB)

    Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

  16. Reactor handbook. 2. rev. ed.

    International Nuclear Information System (INIS)

    Lederer, B.J.; Wildberg, D.W.

    1992-01-01

    On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

  17. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  18. First register of fruit flies (Diptera: Tephritidae in star fruit in Teresina, Altos and Parnaiba, state of Piaui, Brazil/ Primeiro registro de moscas-das-frutas (Diptera: Tephritidae em carambola nos municípios de Teresina, Altos e Parnaíba no estado do Piauí

    Directory of Open Access Journals (Sweden)

    Almerinda Amélia Rodrigues Araújo Soares

    2007-08-01

    Full Text Available The present work aims to register the occurrence of the fruit flies associated to star fruit (Averrhoa carambola L. in three counties of the state of Piaui, as well as to determine the frequency and the index of infestation of these insects. The fruits had been collected during the months of August and September 2005, and had been placed in plastic trays with sterilized soil, stored in metal cages, and left in environmental temperature at the laboratory. Until emergency, the adults had been kept in bottles with alcohol 70% and later identified in the species level. The biggest index of infestation (flies/fruit of C. capitata has occurred in the county of Altos (3.66, followed by Teresina and Parnaiba that had presented index of infestation of 2.18 and 0.016, respectively. C. capitata was the most frequent species in all the counties, presenting frequencies of 100%, 96.5%, and 100% in Teresina, Altos and Parnaiba, respectively. Ceratitis capitata is registered for the first time in star fruit in Teresina, Altos and Parnaiba, state of the Piaui. Anastrepha fraterculus is registered for the first time in the county of Altos. A. fraterculus and C. capitata occur simultaneously in star fruits.O presente trabalho visou conhecer as espécies de moscas-das-frutas associadas à carambola (Averrhoa carambola L. em três municípios do Estado do Piauí, bem como determinar a freqüência e o índice de infestação desses insetos. Os frutos foram coletados durante os meses de agosto e setembro de 2005, colocados em bandejas plásticas com solo esterilizado, armazenados em gaiolas metálicas e deixados em temperatura ambiente no laboratório. Até a emergência dos adultos, estes foram acondicionados em frascos contendo álcool 70% e posteriormente identificados em nível de espécie. O maior índice de infestação (moscas/fruto de C. capitata ocorreu no município de Altos (3,66, seguido pelos municípios de Teresina e Parnaíba que apresentaram

  19. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  20. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, E.

    2007-01-01

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  1. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  2. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  3. Processamento da rede neocognitron para reconhecimento facial em ambiente de alto desempenho GPU

    OpenAIRE

    Gustavo Poli Lameirão da Silva

    2007-01-01

    Neste trabalho é apresentada a implementação da Rede Neural Neocognitron, usando uma arquitetura de computação de alto desempenho baseada em GPU (Graphics Processing Unit). O Neocognitron é uma rede neural artificial, proposta por Fukushima e colaboradores, constituída de vários estágios de camadas de neurônios, organizados em matrizes bidimensionais denominadas planos celulares. Para o processamento de alto desempenho da aplicação de reconhecimento facial usando neocognitron foi utilizado o ...

  4. System Design of a Supercritical CO_2 cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

  5. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  6. Vaccinazione antinfluenzale nella ASL RMF della Regione Lazio: verifica dei risultati e dei costi sostenuti

    Directory of Open Access Journals (Sweden)

    L. Di Marzio

    2003-05-01

    Full Text Available

    Obiettivi: la vaccinazione antinfluenzale nella
    Regione Lazio dalla campagna 1999-2000 viene
    condotta sulla base di un protocollo regionale che,
    per favorire il raggiungimento degli obiettivi stabiliti
    dal Piano Sanitario Nazionale, coinvolge i
    Medici di Medicina Generale (MMG prevedendo
    una remunerazione aggiuntiva in parte fissa (a prestazione, in parte variabile (condizionata dal risultato
    del singolo medico e della ASL.
    Gli autori si propongono una verifica dei risultati raggiunti e dei costi sostenuti dall’ultima campagna eseguita con sole risorse aziendali del 1998-99 a quella del 2002-03.

    Metodi: il protocollo regionale prevede la raccolta
    delle informazioni per ciascun vaccinato presente
    nell’anagrafe informatizzata degli assistiti aziendali
    e ciò consente la valutazione delle coperture vaccinali
    aziendale e per ciascun MMG.
    Parallelamente sono considerati costi dei vaccini
    acquistati e retribuzione aggiuntiva dei MMG.

    Risultati: esaminati gli archivi dal 1998-99 al 2002-
    03, emerge il progressivo coinvolgimento dei MMG fino al recente 97%, l’aumento inequivocabile delle dosi di vaccino somministrate (da 9.406 a 36.692 e del tasso di copertura negli anziani (dal 24,2% al 66%. Invece la percentuale dei vaccini somministrati ai ›65 diminuisce dal 85,47% al 71,77% ed aumenta a favore dei più giovani così da risultare coperture negli ultrasessantacinquenni inferiori alle attese.Con gli anni l’integrazione dell’esperienza del servizio e dei MMG ha favorito un più oculato approvvigionamento
    con diminuzione degli sprechi passando dal 15,56% nel 2000-01 all’attuale 4,45%, ma contestualmente i costi risultano decuplicati (da 90 a 938 milioni di lire per maggior numero di dosi somministrate e costo delle prestazioni dei MMG

  7. Sílica gel obtida de escória de alto forno: Marabá, Pará

    Directory of Open Access Journals (Sweden)

    M. M. Rebelo

    2015-09-01

    Full Text Available ResumoSílica gel com propriedades similares à sílica comercial foi obtida a partir de escória de alto forno (EAF, utilizando digestão com ácido clorídrico. A EAF-sílica obtida foi caracterizada por diferentes técnicas, mostrando-se amorfa, com pureza 99,7% e área específica 282 m2/g. Apresentou caráter hidrofílico alto (12,27%, com água de constituição de ~ 6,18%, o que foi confirmado pela perda de massa durante a análise termogravimétrica. As partículas de EAF-sílica apresentaram tamanhos micrométricos (< 1 µm em forma de agregados, distribuição granulométrica unimodal e D50 7,0 µm.

  8. Technical and scientific report of the Alto project; Rapport scientifique et technique du projet ALTO

    Energy Technology Data Exchange (ETDEWEB)

    Essabaa, S.; Gardes, D.; Grialou, D.; Ibrahim, F.; Le Scornet, J.C

    2002-07-01

    The Alto project means the installation of an electron linear accelerator inside the experimental area of the tandem accelerator of the nuclear physics institute of Orsay (IPNO, France). This linear accelerator comes from CERN where it was operating as a pre-injector for LEP. This equipment will allow IPNO'teams to perform fast kinetics studies in a domain different from that of ELYSE accelerator. The time resolution will not be as high as that of ELYSE (picosecond) but will be sufficient (microsecond) to produce free radicals in aqueous and gaseous media. The main expectations of this installation can be classified according 3 axis: 1) basic research (mainly the study of nuclear matter through photo-fission, 2) research and development of accelerators (by providing a test bench for new high frequency systems and superconducting components), and 3) applied research for industry concerning: biochemistry under irradiation, radiation sensibility, DNA breaking, food and drug sterilization and behaviour of electronic components under irradiation. This rapport details the research program that could be achieved with this equipment, describes its contributions in terms of economic development, cooperation with industry, student training, and specifies the needed investment and the operating and maintenance costs. (A.C.)

  9. Operation of the SLOWPOKE-2 reactor in Jamaica

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

    2001-07-01

    Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

  10. Irradiation techniques at BR2 reactor

    International Nuclear Information System (INIS)

    Hebel, W.

    1978-01-01

    Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

  11. Modelización aplicada al diseño de sistemas de control en el horno alto

    Directory of Open Access Journals (Sweden)

    Rosal, R.

    1995-06-01

    Full Text Available The production of pig iron in blast furnaces resists automatic control strategies due to the lack of knowledge about physical and chemical phenomena taking place inside the reactor. High dimensions lead to important dead times and lags. As a consequence it is very difficult to quantify control actions from actual process measurements. A simplified multizonal mathematical model has been proposed that allowed the description of a given blast furnace excluding hearth. Parameters underlying the model have been identified and, under appropriate assumptions, temperature and composition profiles have been established. The analysis of model predictions has been illustrated with steady-state responses to typical control actions.

    El control del proceso de fabricación de arrabio en hornos altos resulta complejo debido a las condiciones de operación: conocimiento incompleto de la quimicofísica de los procesos que tienen lugar en el interior del homo, grandes dimensiones del reactor que se traducen en tiempos muertos considerables y constantes de tiempo elevadas que provocan una gran inercia a las acciones de control. En este trabajo, se ha planteado un modelo matemático por zonas que permite describir el comportamiento del homo excepto el crisol, se han identificado sus parámetros y se ha obtenido el perfil interno de temperaturas y composiciones. El análisis del modelo permite predecir los efectos de un cambio en cualquier variable del sistema así como desarrollar un algoritmo de control automático.

  12. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  13. TMI-2 reactor vessel plenum final lift

    International Nuclear Information System (INIS)

    Wilson, D.C.

    1986-01-01

    Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

  14. Refurbishment programme for the BR2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

    1992-07-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  15. Refurbishment programme for the BR2-reactor

    International Nuclear Information System (INIS)

    Koonen, E.

    1992-01-01

    BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

  16. El chile poblano criollo en la cultura alimentaria del Alto Atoyac

    Directory of Open Access Journals (Sweden)

    Luis Joaquín Pérez Carrasco

    2017-01-01

    Full Text Available El chile poblano criollo producido en la re-gión Alto Atoyac en Puebla, forma parte de la cultura alimenticia de la población, junto con el maíz y el frijol. Ya sea en fresco o en seco es un componente fundamental en muy diver-sos platillos como: el mole poblano, los chiles en nogada, las rajas con huevo, por mencio-nar algunos. El objetivo del trabajo fue el en-tender las razones sociales y culturales de lo planteado e identificar la problemática del cultivo de chile poblano criollo y los factores que favorecen que los productores persistan en su cultivo en la región. Metodología. Se realizaron entrevistas estructuradas, siguien-do el método de muestreo por “bola de nieve” (Snowball, empleado frecuentemente en es-tudios con poblaciones marginales. Resulta-dos. El sistema de producción predominante en el Alto Atoyac, es el chile poblano criollo intercalado en árboles frutales, con superficies de siembra igual o menor a 100 m2, estrategia usada por los productores para diversificar el riesgo de las enfermedades del cultivo y con ello asegurar la sobrevivencia de sus tradicio-nes culinarias y la permanencia de su semilla con sus propias características. Limitaciones. El trabajo de investigación no pudo abarcar el rendimiento de chile poblano en la región y del perfil del productor. Conclusiones. El chi-le poblano criollo en el Alto Atoyac, se siem-bra en superficies pequeñas y condiciones de temporal, intercalado en árboles frutales y es afectado por la enfermedad pudrición radical o secadera. El productor continúa sembrando su semilla de chile poblano criollo, como estra-tegia para conservar sus tradiciones en la elabo-ración de los alimentos y mitigar en lo posible los daños ocasionados por las enfermedades.

  17. Metallogenic aspects of the feldspars and micas geochemistry in pegmatite from Alto-Ligonha (Mocambique)

    International Nuclear Information System (INIS)

    Neves, J.M.C.

    1990-01-01

    This paper deals with metallogenic aspects concerning the huge Alto Ligonha pegmatite Province. The geological setting of the pegmatites is briefly reviewed and the metamorphic grade of the country rocks of the pegmatites, ranging from granulitic to greenschist facies, has been considered. The economically most interesting pegmatites are those emplaced within rocks with lighter metamorphism. The available geochronological data allow us to link, the most interesting pegmatites from Alto Ligonha, to the Pan-African granitoid magmatism, about 500 Ma ago. (author)

  18. Aesthetic Communication and Intercultural Perspective. A Qualitative Analysis of Aesthetic Perceptions of the Brand "Südtirol/Alto Adige"

    Directory of Open Access Journals (Sweden)

    Vincenzo Bua

    2009-01-01

    Full Text Available In the qualitative study of mental associations with the brand picture "Südtirol/Alto Adige" different images of the region among German speaking, Italian speaking and bilingually grown up South Tyroleans were analysed. The research interest was focused on the communalities and differences in these associations in order to identify potentially conflicting positions between the two major language groups in Südtirol/Alto Adige. In this paper the method is demonstrated which was used to display and investigate the emotional and cognitive contents of the images to Südtirol/Alto Adige from the point of view of different socio-cultural groups. Additionally selected results connected to the perception of the brand in the multilingual province Südtirol/Alto Adige are shown. Against the background of the outlined study the following questions are dealt with in this article: How is the special design of the brand picture perceived among the different socio-cultural groups in Alto Adige/Südtirol with respect to intercultural communication processes? Which meaning can be attributed to the historical heritage of the language groups in the analysis? URN: urn:nbn:de:0114-fqs0901323

  19. Refurbishing the BR2 materials testing reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Dekeyser, J.; Gubel, P.

    1995-01-01

    SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

  20. Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

  1. SIRIUS 2: A versatile medium power research reactor

    International Nuclear Information System (INIS)

    Rousselle, P.

    1992-01-01

    Most of the Research Reactors in the world have been critical in the Sixties and operated for twenty to thirty years. Some of them have been completely shut down, modified, or simply refurbished; the total number of RR in operation has decreased but there is still an important need for medium power research reactors in order: - to sustain a power program with fuel and material testing for NPP or fusion reactors; - to produce radioisotopes for industrial or medical purposes, doped silicon, NAA or neutron radiography; - to investigate further the condensed matter, with cold neutrons routed through neutron guides to improved equipment; - to develop new technologies and applications such as medical alphatherapy. Hence, taking advantage of nearly hundred reactor x years operation and backed up by the CEA experience, TECHNICATOME assisted by FRAMATOME has designed a new versatile multipurpose Research Reactor (20-30 Mw) SIRIUS 2 taking into account: - more stringent safety rules; - the lifetime; - the flexibility enabling a wide range of experiments and, - the future dismantling of the facility according to the ALARA criteria

  2. OTUS - Reactor inventory management system based on ORIGEN2

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

    1995-10-01

    ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

  3. The 2nd reactor core of the NS Otto Hahn

    International Nuclear Information System (INIS)

    Manthey, H.J.; Kracht, H.

    1979-01-01

    Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

  4. The Alto Tiberina Near Fault Observatory (northern Apennines, Italy

    Directory of Open Access Journals (Sweden)

    Lauro Chiaraluce

    2014-06-01

    Full Text Available The availability of multidisciplinary and high-resolution data is a fundamental requirement to understand the physics of earthquakes and faulting. We present the Alto Tiberina Near Fault Observatory (TABOO, a research infrastructure devoted to studying preparatory processes, slow and fast deformation along a fault system located in the upper Tiber Valley (northern Apennines, dominated by a 60 km long low-angle normal fault (Alto Tiberina, ATF active since the Quaternary. TABOO consists of 50 permanent seismic stations covering an area of 120 × 120 km2. The surface seismic stations are equipped with 3-components seismometers, one third of them hosting accelerometers. We instrumented three shallow (250 m boreholes with seismometers, creating a 3-dimensional antenna for studying micro-earthquakes sources (detection threshold is ML 0.5 and detecting transient signals. 24 of these sites are equipped with continuous geodetic GPS, forming two transects across the fault system. Geochemical and electromagnetic stations have been also deployed in the study area. In 36 months TABOO recorded 19,422 events with ML ≤ 3.8 corresponding to 23.36e-04 events per day per squared kilometres; one of the highest seismicity rate value observed in Italy. Seismicity distribution images the geometry of the ATF and its antithetic/synthetic structures located in the hanging-wall. TABOO can allow us to understand the seismogenic potential of the ATF and therefore contribute to the seismic hazard assessment of the area. The collected information on the geometry and deformation style of the fault will be used to elaborate ground shaking scenarios adopting diverse slip distributions and rupture directivity models.

  5. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  6. Physics design of advanced steady-state tokamak reactor A-SSTR2

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

    Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

  7. FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

    International Nuclear Information System (INIS)

    Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

    1984-01-01

    1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

  8. Energy and quality of life: a case study in HPP Tijuco Alto, Ribeira, SP; Energia e qualidade de vida: estudo de caso da Uhe Tijuco Alto no Municipio de Ribeira, SP

    Energy Technology Data Exchange (ETDEWEB)

    Conceicao, Andre Luiz da [Universidade Estadual de Campinas (FEM/UNICAMP), SP (BRazil). Fac. de Engenharia Mecanica], email: conceicao.andreluiz@yahoo.com.br; Seixas, Sonia Regina da Cal [Universidade Estadual de Campinas (NEPAM/UNICAMP), SP (BRazil). Nucleo de Estudos e Pesquisas Ambientais], email: srcal@unicamp.br

    2010-07-01

    This paper deals with a critical and reflexive that the issue involving the possibility of construction and operation of Hydroelectric Power (HEP) Tijuco Alto, the upper course of the Ribeira Valley between Sao Paulo and Parana, in the Vale do Ribeira. This project will directly affect the towns of Ribeira-SP, Itapirapua Paulista-SP, Cerro Azul-PR, Dr. Ulysses-PR and Adrianopolis-PR. Thus, we defined the main objective of the research examines the quality of life in the city of Ribeira-SP, at the possibility of deployment of the dam. Thus, field research was conducted in the city and interviews with residents, where it was possible to observe, among other things the precarious economic conditions, social, urban and cultural community. Another aspect noted was the fact that most respondents to position themselves for the construction of the HPP Tijuco Alto, citing primarily the need for local development and increased job opportunities. Those opposing the plant, highlighted environmental issues, mainly, reasons related to loss of peace and security site. Regardless of those who are for or against, a technical opinion issued by IBAMA in 2008 points to the likely deployment of the HPP Tijuco Alto. (author)

  9. Motivación y equipos de alto desempeño

    OpenAIRE

    Olaya Gómez, Audrey Yazmin

    2015-01-01

    “Yo hago lo que usted no puede, y usted hace lo que yo no puedo. Juntos podemos hacer grandes cosas” (Teresa de Calcuta).El objetivo de este ensayo es reconocer que en un equipo de alto desempeño se requiere tanta motivación a nivel personal. 

  10. Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

    International Nuclear Information System (INIS)

    Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

    2010-01-01

    The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

  11. PANIC: A General-purpose Panoramic Near-infrared Camera for the Calar Alto Observatory

    Science.gov (United States)

    Cárdenas Vázquez, M.-C.; Dorner, B.; Huber, A.; Sánchez-Blanco, E.; Alter, M.; Rodríguez Gómez, J. F.; Bizenberger, P.; Naranjo, V.; Ibáñez Mengual, J.-M.; Panduro, J.; García Segura, A. J.; Mall, U.; Fernández, M.; Laun, W.; Ferro Rodríguez, I. M.; Helmling, J.; Terrón, V.; Meisenheimer, K.; Fried, J. W.; Mathar, R. J.; Baumeister, H.; Rohloff, R.-R.; Storz, C.; Verdes-Montenegro, L.; Bouy, H.; Ubierna, M.; Fopp, P.; Funke, B.

    2018-02-01

    PANIC7 is the new PAnoramic Near-Infrared Camera for Calar Alto and is a project jointly developed by the MPIA in Heidelberg, Germany, and the IAA in Granada, Spain, for the German-Spanish Astronomical Center at Calar Alto Observatory (CAHA; Almería, Spain). This new instrument works with the 2.2 m and 3.5 m CAHA telescopes covering a field of view of 30 × 30 arcmin and 15 × 15 arcmin, respectively, with a sampling of 4096 × 4096 pixels. It is designed for the spectral bands from Z to K S , and can also be equipped with narrowband filters. The instrument was delivered to the observatory in 2014 October and was commissioned at both telescopes between 2014 November and 2015 June. Science verification at the 2.2 m telescope was carried out during the second semester of 2015 and the instrument is now at full operation. We describe the design, assembly, integration, and verification process, the final laboratory tests and the PANIC instrument performance. We also present first-light data obtained during the commissioning and preliminary results of the scientific verification. The final optical model and the theoretical performance of the camera were updated according to the as-built data. The laboratory tests were made with a star simulator. Finally, the commissioning phase was done at both telescopes to validate the camera real performance on sky. The final laboratory test confirmed the expected camera performances, complying with the scientific requirements. The commissioning phase on sky has been accomplished.

  12. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

    1983-02-01

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  13. Thermal reactor benchmark tests on JENDL-2

    International Nuclear Information System (INIS)

    Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

    1983-11-01

    A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

  14. Identidades en movimiento: familias chilenas en la fruticultura del Alto Valle de Río Negro, Argentina Identities in movement: chilean families in the fruit production of the Alto Valle de Río Negro, Argentina

    Directory of Open Access Journals (Sweden)

    Verónica Trpin

    2007-12-01

    Full Text Available Este artículo, basado en el trabajo de campo realizado en áreas rurales del Alto Valle de Río Negro, Argentina, desde el año 1999, tiene como propósito presentar las relaciones en las cuales se insertan hombres y mujeres chilenas que residen y trabajan en "chacras" destinadas a la producción frutícola. Las diferentes actividades en las chacras se organizan según el sexo y la edad, definiéndose una segmentación del mercado de trabajo en la que se ven involucrados los diferentes miembros de la familia. Como desarrollaré, ser trabajadores chilenos en la fruticultura del Alto Valle de Río Negro reproduce una identidad étnica y nacional en el seno de la cotidianeidad familiar y laboral.This article, based on field work conducted in rural areas of the Alto Valle de Río Negro, Argentina, from 1999 on, analyzes the relations in which Chilean men and women who reside and work in small farms destined to fruit production are inserted. The different activities in the small farms are organized according to sex and age, circumscribing a segment of the labor market in which different members of the family are involved. As I will demonstrate, to be a Chilean worker in the fruit growing region of the Alto Valle is to reproduce an ethnic and national identity through work routines mediated by family relations.

  15. Sonido espacial para una inmersión audiovisual de alto realismo

    Directory of Open Access Journals (Sweden)

    Basilio Pueo Ortega

    2012-04-01

    Full Text Available Los sistemas de vídeo y audio de alta inmersión tienen un auge impor-tante en entornos audiovisuales realistas. Las sensaciones visuales y sonoras que crean en el público se aproximan con un alto grado de similitud a lo percibido en el entorno real que pretenden recrear. Para ello, los estímulos deben contener toda la información necesaria, tanto espacial como temporal, que permita crear la ilusión de que el objeto audiovisual es real. En este artículo, se realiza un repaso de los sistemas audiovisuales que permiten esta recreación, con especial atención en los sistemas de audio envolvente. Se describe la técnica de audio 3D más prometedora, Wave Field Synthesis, junto con diversos campos de aplicación de entornos audiovisuales de alto realismo.

  16. Angioplastia del tronco de la arteria coronaria izquierda no protegido en pacientes con alto riesgo quirúrgico

    Directory of Open Access Journals (Sweden)

    Victor Mauro

    2008-01-01

    Full Text Available IntroducciónEl tratamiento de elección de la enfermedad del tronco de la coronaria izquierda (TCI es la cirugía de revascularización miocárdica (CRM. Un número creciente de pacientes presenta comorbilidades y/o inestabilidad clínica que condicionan un alto riesgo quirúrgico.ObjetivosEvaluar los resultados de la angioplastia (ATC del TCI no protegido en pacientes con alto riesgo para CRM (EUROSCORE = 6.Material y métodosDe 59 pacientes con ATC de TCI no protegido se excluyeron 8 con infarto agudo de miocardio (IAM en shock cardiogénico y 12 sin características de alto riesgo; de los restantes pacientes de alto riesgo fueron objeto de este estudio los 32 tratados con stents convencionales.Se comparó la mortalidad hospitalaria predicha por EUROSCORE logístico con la observada, así como la incidencia de complicaciones mayores y su evolución alejada.ResultadosLa mediana de edad fue de 76,5 años, el 41% tenía 80 años o más, el 22% eran mujeres, el 28% diabéticos, el 56% tenía disfunción ventricular moderada a grave, el 31% insuficiencia renal crónica, el 50% vasculopatía periférica, el 53% angina refractaria, el 22% IAM reciente, el 28% procedimientos de emergencia y la mediana de EUROSCORE fue de 10,5 puntos.El 41% de los pacientes presentaban compromiso del TCI distal. El éxito angiográfico fue del 94%. Se utilizaron inhibidores IIb/IIIa en el 47%, cutting balloon en el 28%, Rotablator® en el 3% y balón de contrapulsación en el 31%. En todos se implantó un stent y en el 50% se trataron otras obstrucciones.La mortalidad hospitalaria fue del 3,1% (intervalo de confianza del 95% 0,2%-14,5%, p = 0,003, en tanto que la predicha era del 23,8%. Ningún paciente presentó déficit neurológico, IAM transmural ni requirió diálisis. Un paciente debió ser sometido a CRM electiva por fracaso del procedimiento.La mediana de seguimiento fue de 15,5 meses, período en el que se registraron 6 muertes (2 cardiovasculares y 4

  17. WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

  18. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  19. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

    2014-01-01

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  20. Prescrição de terapias baseadas em evidências para pacientes de alto risco cardiovascular: estudo REACT

    Directory of Open Access Journals (Sweden)

    Otávio Berwanger

    2013-03-01

    Full Text Available FUNDAMENTO: Dados de atendimento ambulatorial ao paciente de alto risco cardiovascular no Brasil são insuficientes. OBJETIVO: Descrever o perfil e documentar a prática clínica do atendimento ambulatorial de pacientes de alto risco cardiovascular no Brasil, no que diz respeito à prescrição de terapias baseadas em evidências. MÉTODOS: Registro prospectivo que documentou a prática clínica ambulatorial de indivíduos de alto risco cardiovascular, que foi definido como a presença de um dos seguintes fatores: doença arterial coronariana, cerebrovascular e vascular periférica; diabetes; ou aqueles com pelo menos três dos seguintes fatores: hipertensão arterial, tabagismo, dislipidemia, maiores 70 anos, histórico familiar de doença arterial coronariana, nefropatia crônica ou doença carotídea assintomática. Foram avaliadas características basais e a taxa de prescrição das intervenções medicamentosas e não medicamentosas. RESULTADOS: Foram incluídos 2.364 pacientes consecutivos, sendo 52,2% do gênero masculino, idade média de 66,0 anos (± 10,1. Dentre os pacientes incluídos, 78,3% utilizavam antiplaquetários, 77,0% estatinas e, dos pacientes com história de infarto do miocárdio, 58,0% receberam betabloqueadores. O uso concomitante destas três classes foi de 34%. Não atingiram as metas preconizadas pelas diretrizes 50,9% dos hipertensos, 67% dos diabéticos e 25,7% dos dislipidêmicos. Os principais preditores de prescrição de terapias com benefício comprovado foram centro com cardiologista e histórico de doença arterial coronariana. CONCLUSÃO: Este registro nacional e representativo identificou hiatos importantes na incorporação de terapias com benefício comprovado, oferecendo um panorama real dos pacientes de alto risco cardiovascular.

  1. Variación antropométrica y nutricional en Susques y Alto Comedero entre 2002-2007

    Directory of Open Access Journals (Sweden)

    Bejarano, Ignacio

    2007-01-01

    Full Text Available Las poblaciones humanas experimentan variaciones de los parámetros antropométricos como expresión de los cambios socioambientales. El objetivo de este trabajo fue analizar la variación temporal de talla, peso y estado nutricional en dos poblaciones jujeñas situadas a distintos niveles altitudinales. Los datos procedieron de mediciones realizadas en 2002 y 2007 en poblaciones de 6 a 17 años de Susques (3500 m y Alto Comedero (1200 m. Se calcularon las categorías nutricionales de Waterlow y las diferencias entre talla y peso y categorías nutricionales se establecieron con ANOVA y prueba de comparación de proporciones (χ2 respectivamente. Para ambas poblaciones se observaron diferencias interanuales estadísticamente significativas de los promedios de talla y peso, siendo menores en Susques en el 2007, lo contrario sucede en Alto Comedero. Las diferencias interanuales de la categorías nutricionales no fueron estadísticamente significativas en Alto Comedero, pero si en Susques para normonutridos y obesos que disminuyeron y aumentaron respectivamente entre 2002 y 2007. En el contexto de las modificaciones socioeconómicas experimentadas por la población susqueña en los últimos años, debido a su mayor conexión e integración con poblaciones vecinas por la apertura del Paso de Jama, los resultados indicarían un empeoramiento de las condiciones nutricionales de su población infanto juvenil.

  2. CALIFA, the Calar Alto Legacy Integral Field Area survey. IV. Third public data release

    Science.gov (United States)

    Sánchez, S. F.; García-Benito, R.; Zibetti, S.; Walcher, C. J.; Husemann, B.; Mendoza, M. A.; Galbany, L.; Falcón-Barroso, J.; Mast, D.; Aceituno, J.; Aguerri, J. A. L.; Alves, J.; Amorim, A. L.; Ascasibar, Y.; Barrado-Navascues, D.; Barrera-Ballesteros, J.; Bekeraitè, S.; Bland-Hawthorn, J.; Cano Díaz, M.; Cid Fernandes, R.; Cavichia, O.; Cortijo, C.; Dannerbauer, H.; Demleitner, M.; Díaz, A.; Dettmar, R. J.; de Lorenzo-Cáceres, A.; del Olmo, A.; Galazzi, A.; García-Lorenzo, B.; Gil de Paz, A.; González Delgado, R.; Holmes, L.; Iglésias-Páramo, J.; Kehrig, C.; Kelz, A.; Kennicutt, R. C.; Kleemann, B.; Lacerda, E. A. D.; López Fernández, R.; López Sánchez, A. R.; Lyubenova, M.; Marino, R.; Márquez, I.; Mendez-Abreu, J.; Mollá, M.; Monreal-Ibero, A.; Ortega Minakata, R.; Torres-Papaqui, J. P.; Pérez, E.; Rosales-Ortega, F. F.; Roth, M. M.; Sánchez-Blázquez, P.; Schilling, U.; Spekkens, K.; Vale Asari, N.; van den Bosch, R. C. E.; van de Ven, G.; Vilchez, J. M.; Wild, V.; Wisotzki, L.; Yıldırım, A.; Ziegler, B.

    2016-10-01

    This paper describes the third public data release (DR3) of the Calar Alto Legacy Integral Field Area (CALIFA) survey. Science-grade quality data for 667 galaxies are made public, including the 200 galaxies of the second public data release (DR2). Data were obtained with the integral-field spectrograph PMAS/PPak mounted on the 3.5 m telescope at the Calar Alto Observatory. Three different spectral setups are available: I) a low-resolution V500 setup covering the wavelength range 3745-7500 Å (4240-7140 Å unvignetted) with a spectral resolution of 6.0 Å (FWHM) for 646 galaxies, II) a medium-resolution V1200 setup covering the wavelength range 3650-4840 Å (3650-4620 Å unvignetted) with a spectral resolution of 2.3 Å (FWHM) for 484 galaxies, and III) the combination of the cubes from both setups (called COMBO) with a spectral resolution of 6.0 Å and a wavelength range between 3700-7500 Å (3700-7140 Å unvignetted) for 446 galaxies. The Main Sample, selected and observed according to the CALIFA survey strategy covers a redshift range between 0.005 and 0.03, spans the color-magnitude diagram and probes a wide range of stellar masses, ionization conditions, and morphological types. The Extension Sample covers several types of galaxies that are rare in the overall galaxy population and are therefore not numerous or absent in the CALIFA Main Sample. All the cubes in the data release were processed using the latest pipeline, which includes improved versions of the calibration frames and an even further improved image reconstruction quality. In total, the third data release contains 1576 datacubes, including ~1.5 million independent spectra. Based on observations collected at the Centro Astronómico Hispano Alemán (CAHA) at Calar Alto, operated jointly by the Max-Planck-Institut für Astronomie (MPIA) and the Instituto de Astrofísica de Andalucía (CSIC).The spectra are available at http://califa.caha.es/DR3

  3. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  4. Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

    International Nuclear Information System (INIS)

    Villeger, S; Cousty, S; Ricard, A; Sixou, M

    2003-01-01

    Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

  5. Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

    International Nuclear Information System (INIS)

    ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

    2017-01-01

    The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

  6. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  7. ¿El buen entrenador nace o lo hace el deportista? El camino hacia el alto nivel en triatlón

    Directory of Open Access Journals (Sweden)

    Germ\\u00E1n Ruiz Tendero

    2014-01-01

    Full Text Available La figura del entrenador adquiere un peso importante en el sistema deportivo y por tanto en el éxito de sus deportistas. Las claves de su éxito han sido estudiadas desde diferentes perspectivas. El estudio en retrospectiva del recorrido por el cual se llega al alto nivel es una de ellas. El propósito de este estudio fue determinar el camino de los entrenadores de triatlón previo a su llegada al alto nivel, así como las circunstancias en las que se produjo el paso hacia el alto rendimiento. Para ello se entrevistó a una muestra de 14 entrenadores españoles de alto nivel en triatlón. Los resultados muestran un recorrido prevalente en el que el entrenador fue anteriormente deportista y entrenador en alguna/s de las disciplinas fundamentales (DF de las que se compone el triatlón (natación, ciclismo, atletismo, llegando al alto nivel de triatlón con una edad aproximada de 30 años. Los años de experiencia previa varían en función del pasado del entrenador, no llegándose a alcanzar los 10 años de media en ningún caso, hasta el inicio en la etapa de alto nivel. Sería recomendable, por tanto, contextualizar los años de experiencia previos, para optimizar la selección de muestras de entrenadores expertos.

  8. PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

  9. Valoración del deporte de alto rendimiento (gimnasia rítmica) en edades tempranas

    OpenAIRE

    Usero Gómez, Alba

    2014-01-01

    Comenzando por una introducción, en la cual se contextualiza el deporte y especialmente el de alto rendimiento, nos introduciremos en la cuestión de estudio, el deportista de élite y la preocupación por el comienzo en edades tempranas. Llevaremos a cabo este estudio, por medio de un análisis reflexivo de diversos autores y estudios que se sumergen en el deporte de alto rendimiento, especialmente en la infancia. Trataremos el objeto de estudio en relación a un deporte, la Gimnasia Rítmic...

  10. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  11. El 'ayllu' reterritorializado, y su 'taypi'. La ciudad de El Alto.

    Directory of Open Access Journals (Sweden)

    Orlando Augusto Yépez Mariaca

    2010-05-01

    Twenty-five years a suburb of La Paz, now the city of El Alto, the heat of the capital's neoliberal policies, implodes in the urban area provided by flat topography as opposed to La Paz, to become today in a city with larger population and greater extent than its parent. With a population of mostly Aymara-Indian-moving and rich in its live, old traditions of the Andean Community institution like Ayllu and Aini, among others. On October 2003, the city of El Alto, the epicenter of a massive social upheaval, becoming the leader of the anti-globalization social movements. Will the 'pachakuti' "return" of the ancient traditions originate? The shop is above all 'live together', and perhaps a light at the end of the tunnel, a tunnel that big business has been built so arrogant and conceited, leaving cities now fragmented, unbalanced territories and a planet on the brink of collapse.

  12. Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.

    1991-11-01

    This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

  13. Shadow corrosion evaluation in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Sanders, Ch.; Lysell, G.

    2000-01-01

    Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

  14. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  15. Radionuclide distribution in TMI-2 reactor building basement liquids and solids

    International Nuclear Information System (INIS)

    Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

    1984-01-01

    As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

  16. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  17. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    International Nuclear Information System (INIS)

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

  18. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  19. Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

    International Nuclear Information System (INIS)

    Chatt, A.; Holzbecher, J.

    1990-01-01

    SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

  20. Technical and scientific report of the Alto project

    CERN Document Server

    Essabaa, S; Grialou, D; Ibrahim, F; Le Scornet, J C

    2002-01-01

    The Alto project means the installation of an electron linear accelerator inside the experimental area of the tandem accelerator of the nuclear physics institute of Orsay (IPNO, France). This linear accelerator comes from CERN where it was operating as a pre-injector for LEP. This equipment will allow IPNO'teams to perform fast kinetics studies in a domain different from that of ELYSE accelerator. The time resolution will not be as high as that of ELYSE (picosecond) but will be sufficient (microsecond) to produce free radicals in aqueous and gaseous media. The main expectations of this installation can be classified according 3 axis: 1) basic research (mainly the study of nuclear matter through photo-fission, 2) research and development of accelerators (by providing a test bench for new high frequency systems and superconducting components), and 3) applied research for industry concerning: biochemistry under irradiation, radiation sensibility, DNA breaking, food and drug sterilization and behaviour of electro...

  1. CALIFA, the Calar alto legacy integral field area survey

    DEFF Research Database (Denmark)

    Husemann, B.; Jahnke, K.; Sánchez, S. F.

    2013-01-01

    We present the first public data release (DR1) of the Calar Alto Legacy Integral Field Area (CALIFA) survey. It consists of science-grade optical datacubes for the first 100 of eventually 600 nearby (0.005 < z < 0.03) galaxies, obtained with the integral-field spectrograph PMAS/PPak mounted on th...... the available interfaces and tools that allow easy access to this first publicCALIFA data at http://califa.caha.es/DR1....

  2. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

  3. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  4. Dynamic simulation of the 2 MWt slowpoke heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Lepp, R.M.

    1982-04-01

    A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

  5. Distribution of energy of impulses of the modernized IBR-2 REACTOR

    International Nuclear Information System (INIS)

    Tayibov, L.A; Mehtiyeva, R.N.; )

    2011-01-01

    Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

  6. Los tambos Inca: el caso de Camata Tambo valle alto de Moquegua

    OpenAIRE

    Chacaltana Cortez, Sofía; Ministerio de Cultura

    2013-01-01

    Camata Tambo está ubicado en la parte alta del valle alto de Moquegua. Por este tambo pasa un camino Inca que viene del altiplano y continúa hacia el centro provincial de Sabaya ubicado a 1 km valle abajo.

  7. Equipment for thermal neutron flux measurements in reactor R2

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Nilsson, T; Claeson, S

    1960-04-15

    For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

  8. A conceptual design of LIB fusion reactor: UTLIF(2)

    International Nuclear Information System (INIS)

    Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

    1984-01-01

    UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

  9. Inhaled Corticosteroid Use in Chronic Obstructive Pulmonary Disease and Risk of Pneumonia: A Nested Case-Control Population-based Study in Lazio (Italy)-The OUTPUL Study.

    Science.gov (United States)

    Cascini, Silvia; Kirchmayer, Ursula; Belleudi, Valeria; Bauleo, Lisa; Pistelli, Riccardo; Di Martino, Mirko; Formoso, Giulio; Davoli, Marina; Agabiti, Nera

    2017-06-01

    Inhaled corticosteroid (ICS) use in chronic obstructive pulmonary disease (COPD) patients is associated with a reduction of exacerbations and a potential risk of pneumonia. The objective was to determine if ICS use, with or without long-acting β 2 -agonist, increases pneumonia risk in COPD patients. A cohort study was performed using linked hospital and drug prescription databases in the Lazio region. Patients (45+) discharged with COPD in 2006-2009 were enrolled and followed from cohort entry until first admission for pneumonia, death or study end, 31 December, 2012. A nested case-control approach was used to estimate the rate ratio (RR) associated with current or past use of ICS adjusted for age, gender, number of exacerbations in the previous year and co-morbidities. Current users were defined as patients with their last ICS prescribed in the 60 days prior to the event. Past users were those with the last prescription between 61 and 365 days before the event. Current use was classified into three levels (high, medium, low) according to the medication possession ratio. Among the cohort of 19288 patients, 3141 had an event of pneumonia (incidence rate for current use 87/1000py, past use 32/1000py). After adjustment, patients with current use were 2.29 (95% confidence interval [CI]: 1.99-2.63) times more likely to be hospitalised for pneumonia with respect to no use; for past use RR was 1.23 (95% CI: 1.07-1.42). For older patients (80+), the rate was higher than that for younger patients. ICS use was associated with an excess risk of pneumonia. The effect was greatest for higher doses and in the very elderly.

  10. The BR2 materials testing reactor. Past, ongoing and under-study upgradings

    Energy Technology Data Exchange (ETDEWEB)

    Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  11. TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

    Directory of Open Access Journals (Sweden)

    Yasmine Abdel-Maksoud

    2016-09-01

    Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

  12. Índice hiperbárico en la prevención primaria de las complicaciones hipertensivas del embarazo de alto riesgo

    Directory of Open Access Journals (Sweden)

    Alfonso Otero González

    2015-11-01

    Resultados: En las gestantes remitidas a consulta de Nefrología (38,2%, diagnosticadas de alto riesgo de PE y tratadas con AAS 100 mg nocturno (desde la semana 17 se redujo la incidencia de episodios de PE un 96,94%.

  13. Planned Scientific programs around the Triga Mark 2 Reactor

    International Nuclear Information System (INIS)

    Majah, M Ibn.

    2007-01-01

    Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

  14. Metodologia de avaliação e desenvolvimento de grupos de alto desempenho

    Directory of Open Access Journals (Sweden)

    Ana Cristina Carneiro

    2008-11-01

    Full Text Available Este artigo discute a fundamentação teórica do Projeto de Avaliação e Desenvolvimento de Grupos de Alto Desempenho, concebido com base na metodologia da Meta-aprendizagem, e no Modelo Evolutivo, estendido à luz da Teoria da Complexidade. Visa ao desenvolvimento e aplicação de uma metodologia de avaliação/constituição de grupos de alto desempenho no ambiente de pesquisa e pós-graduação. A metodologia proposta validada empiricamente teve base no aproveitamento das virtudes e potencialidades das teorias que lhe deram origem. É destinado aos docentes e pesquisadores de vários campos do conhecimento, bem como aos dirigentes de instituições de educação superior e de pesquisa.

  15. Power noise spectrum classification in the problem of the IBR-2 reactor

    International Nuclear Information System (INIS)

    Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

    1988-01-01

    The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

  16. Research reactor FR2 - 20 years chemical and radiochemical measurements

    International Nuclear Information System (INIS)

    Feuerstein, H.; Graebner, H.; Oschinski, J.; Hoffmann, W.; Beyer, J.

    1986-09-01

    The FR2 has been a D 2 O cooled and moderated research reactor with a thermal output of 44 MW. It was in operation from 1961 to 1981. Because of the operating conditions of the reactor, only a small number of routine measurements were performed. For these however special techniques had to be developed. During the 20 years of operation a number of special events occured or have been observed, sometimes with very amazing results, e.g. the 'aceton effect'. This report describes the chemical and radiochemical conditions of the reactor systems, as well as the results of the surveilance work. Not described are measurements for the many experiments. The last chapter gives in a short form a description of the most unusual events and observations. (orig.) [de

  17. Venting krypton-85 from the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Burton, H.M.

    1981-01-01

    To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

  18. Research on economics and CO2 emission of magnetic and inertial fusion reactors

    International Nuclear Information System (INIS)

    Mori, Kenjiro; Yamazaki, Kozo; Oishi, Tetsutarou; Arimoto, Hideki; Shoji, Tatsuo

    2011-01-01

    An economical and environment-friendly fusion reactor system is needed for the realization of attractive power plants. Comparative system studies have been done for magnetic fusion energy (MFE) reactors, and been extended to include inertial fusion energy (IFE) reactors by Physics Engineering Cost (PEC) system code. In this study, we have evaluated both tokamak reactor (TR) and IFE reactor (IR). We clarify new scaling formulas for cost of electricity (COE) and CO 2 emission rate with respect to key design parameters. By the scaling formulas, it is clarified that the plant availability and operation year dependences are especially dominant for COE. On the other hand, the parameter dependences of CO 2 emission rate is rather weak than that of COE. This is because CO 2 emission percentage from manufacturing the fusion island is lower than COE percentage from that. Furthermore, the parameters dependences for IR are rather weak than those for TR. Because the CO 2 emission rate from manufacturing the laser system to be exchanged is very large in comparison with CO 2 emission rate from TR blanket exchanges. (author)

  19. Development of a TiO2-coated optical fiber reactor for water decontamination

    International Nuclear Information System (INIS)

    Danion, A.

    2004-09-01

    The objective of this study was to built and to study a photo-reactor composed by TiO 2 -coated optical fibers for water decontamination. The physico-chemical characteristics and the optical properties of the TiO 2 coating were first studied. Then, the influences of different parameters as the coating thickness, the coating length and the coating volume were investigated both on the light transmission in the TiO 2 - coated fiber and on the photo-catalytic activity of the fiber for a model compound (malic acid). The photo-catalytic degradation of malic acid was optimized using the experimental design methodology allowing to build a multi-fiber reactor comprising 57 optical fibers. The photo-degradation of malic acid was conducted in the multi-fiber reactor and it was demonstrated that the multi-fiber reactor was more efficient than the single-fiber reactor at the same fibers density. Finally, the multi-fiber reactor was applied to the photo-degradation of a fungicide, called fenamidone, and a degradation pathway was proposed. (author)

  20. TMI-2 reactor-vessel head removal and damaged-core-removal planning

    International Nuclear Information System (INIS)

    Logan, J.A.; Hultman, C.W.; Lewis, T.J.

    1982-01-01

    A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

  1. The Alto Paraguay Alkaline Province: petrographic, geochemical and geochronological characteristics; Provincia alcalina Alto Paraguai: caracteristicas petrograficas, geoquimicas e geocronologicas

    Energy Technology Data Exchange (ETDEWEB)

    Velazquez Fernandez, Victor

    1996-12-31

    The Alto Paraguay Province is located at the border of the State of Mato Grosso do Sul and Paraguay, between the coordinates 21 deg 10{sup `}to 23 deg 25{sup `}of Southern latitude and 57 deg 10{sup `} to 58 deg 00{sup `}, having the city of Porto Murtinho as the main reference point. The geotectonic domain of the area is governed by the precambric units of the Southern extreme of the Amazonic craton which developed a long and accentuated activity, giving rise to folds and important faults, that in several cases seem to have exerted an effective control of the magmatic manifestations. Radiometric data indicate that the emplacement of the syenitic bodies took place in the Permo-Triassic period, with a major incidence in the interval 260-240 Ma, representing thus, an important phase of alkaline magmatic affinity associated to the Parana Basin which is believed is to be unique, since the other known areas (Central, Amambay and Rio Apa Provinces, Paraguay, Velasco Province, Bolivia) are considerably younger (140-120 Ma). Syenitic rocks from the Alto Paraguay Province show wide variation in the ratio {sup 87} Sr/{sup 86} Sr (0.703361 - 0.707734). Excluding the Cerro Boggiani rocks (0.703837-0.707734), values for the nepheline syenites (0.703361-0.703672) general lower than those of the other syenites types. Alkaline syenites cover the interval 0.703510- 0.703872, while quartz syenites and syenogranites are 0.704562 and 0.707076, respectively. geologic evidence, in addition to petrographic, geochemical and isotopic (Sr) data, suggest that the syenitic rocks have been derived from an unique mantelic parental liquid, by fractional crystallization and assimilation processes, which are assumed to be occurred during the emplacement of the magma in the crust. (author) 124 refs., 52 figs., 7 tabs.

  2. The Alto Paraguay Alkaline Province: petrographic, geochemical and geochronological characteristics; Provincia alcalina Alto Paraguai: caracteristicas petrograficas, geoquimicas e geocronologicas

    Energy Technology Data Exchange (ETDEWEB)

    Velazquez Fernandez, Victor

    1997-12-31

    The Alto Paraguay Province is located at the border of the State of Mato Grosso do Sul and Paraguay, between the coordinates 21 deg 10{sup `}to 23 deg 25{sup `}of Southern latitude and 57 deg 10{sup `} to 58 deg 00{sup `}, having the city of Porto Murtinho as the main reference point. The geotectonic domain of the area is governed by the precambric units of the Southern extreme of the Amazonic craton which developed a long and accentuated activity, giving rise to folds and important faults, that in several cases seem to have exerted an effective control of the magmatic manifestations. Radiometric data indicate that the emplacement of the syenitic bodies took place in the Permo-Triassic period, with a major incidence in the interval 260-240 Ma, representing thus, an important phase of alkaline magmatic affinity associated to the Parana Basin which is believed is to be unique, since the other known areas (Central, Amambay and Rio Apa Provinces, Paraguay, Velasco Province, Bolivia) are considerably younger (140-120 Ma). Syenitic rocks from the Alto Paraguay Province show wide variation in the ratio {sup 87} Sr/{sup 86} Sr (0.703361 - 0.707734). Excluding the Cerro Boggiani rocks (0.703837-0.707734), values for the nepheline syenites (0.703361-0.703672) general lower than those of the other syenites types. Alkaline syenites cover the interval 0.703510- 0.703872, while quartz syenites and syenogranites are 0.704562 and 0.707076, respectively. geologic evidence, in addition to petrographic, geochemical and isotopic (Sr) data, suggest that the syenitic rocks have been derived from an unique mantelic parental liquid, by fractional crystallization and assimilation processes, which are assumed to be occurred during the emplacement of the magma in the crust. (author) 124 refs., 52 figs., 7 tabs.

  3. Otium, materialidade e paisagem nas villae do Alto Alentejo português em época romana = Otium, Materiality and Landscape in the Roman Villae of Alto Alentejo (Portugal

    Directory of Open Access Journals (Sweden)

    André Carneiro

    2015-03-01

    Full Text Available A arquitectura das villae foi cuidadosamente pensada para permitir o máximo desfrute de uma vivência de gosto urbano e cosmopolita. A atenção dada à inserção da construção na paisagem, as soluções para harmonizar o espaço exterior criando atmosferas favoráveis, a contemplação para o exterior e a criação de espaços e ambientes construídos que permitissem potenciar o otium e o convivium são discutidos neste trabalho, com exemplos de sítios no Alto Alentejo.Roman villae were carefully designed to fulfil the urban and cosmopolitan way of living. Considering some archaeological sites in Alto Alentejo (Portugal, one intends to discuss the adjustment of the built structure to the landscape, the creation of chosen atmospheres by modelling the outer space, the countryside contemplation and the creation of spaces and indoor environments that would promote otium and convivium.

  4. Saúde reprodutiva e mulheres indígenas do Alto Rio Negro

    Directory of Open Access Journals (Sweden)

    Marta Azevedo

    Full Text Available O presente artigo descreve e analisa as concepções próprias das mulheres indígenas do Alto Rio Negro sobre saúde reprodutiva, relacionando-as a indicadores de fecundidade. As informações qualitativas apontam para um conhecimento detalhado e complexo que as mulheres indígenas dessa região possuem sobre seu corpo e os cuidados com sua saúde. Os níveis e padrões etários da fecundidade estão relacionados com a etnia das mulheres, portanto, aos sistemas tradicionais de cuidados com a saúde desses povos. A pesquisa foi desenvolvida entre 1997 e 2003, na região de Iauaretê, Terra Indígena Alto Rio Negro (AM, e teve como primeira fonte de dados o Censo Indígena Autônomo do Rio Negro - CIARN-, levado a efeito pela Federação das Organizações Indígenas do Rio Negro - FOIRN - em 1992.

  5. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  6. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    Ipek, O.

    1997-01-01

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  7. A theoretical analysis of methanol synthesis from CO2 and H2 in a ceramic membrane reactor

    NARCIS (Netherlands)

    Gallucci, F.; Basile, A.

    2007-01-01

    In this theoretical work the CO2 conversion into methanol in both a traditional reactor (TR) and a membrane reactor (MR) is considered. The purpose of this study was to investigate the possibility of increasing CO2 conversion into methanol with respect to a TR. A zeolite MR, able to combine

  8. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  9. Adsorción de metales pesados sobre lodos de horno alto

    Directory of Open Access Journals (Sweden)

    López-Delgado, A.

    1998-05-01

    Full Text Available Most of industrial liquid effluents have high contents of heavy metals. The recovery of these metals is environmental and economically interesting. In this work we study the use of sludge, a by-product of the steel industry, as an adsorbent for the removal of heavy metals from liquid effluents. The adsorption of Pb2+, Zn2+, Cd2+, Cu2+ and Cr3+ on the sludge was investigated by determination of adsorption isotherms. The effect of time, equilibrium temperature and concentration of metal solution on sludge adsorption efficiency was evaluated. The adsorption process was analysed using the theories of Freundlich and Langmuir and the thermodynamic values ΔG, ΔH and ΔS corresponding to each adsorption process were calculated. Blast furnace sludge was found to be an effective sorbent for Pb, Zn, Cd, Cu and Cr-ions within the range of ion concentrations employed.

    Los efluentes líquidos de la mayoría de los procesos industriales contienen una carga importante de metales pesados que, por motivos tanto económicos como medioambientales interesa recuperar. En este trabajo, se estudia la utilización de lodos procedentes de la depuración por vía húmeda de los gases de horno alto como soporte para la retención de metales pesados contenidos en efluentes líquidos. La adsorción de Pb2+, Zn2+, Cd2+, Cu2+ y Cr3+ sobre el lodo de horno alto se determina mediante la obtención de las isotermas de adsorción, variando la concentración de las soluciones metálicas y analizando la influencia del tiempo y de la temperatura de equilibrio en la capacidad de adsorción del lodo. Para describir el proceso de adsorción se consideraron las teorías de Freundlich y Langmuir y, posteriormente, se calcularon los valores termodinámicos ΔG, ΔH y ΔS, correspondientes a cada proceso de adsorción. Para todos

  10. Changes without changes: the Puebla's Alto Atoyac sub-basin case in Mexico

    NARCIS (Netherlands)

    Bressers, Johannes T.A.; Casiano Flores, Cesar Augusto

    2015-01-01

    Since the year 2000, actions at the three governmental levels have taken place to improve water quality in Mexico’s Puebla Alto Atoyac sub-basin. This paper reports a situation in which several policy actors have been striving for water quality improvement in that polluted sub-basin. However, when

  11. Schooling and Critical Citizenship: Pedagogies of Political Agency in El Alto, Bolivia

    Science.gov (United States)

    Lazar, Sian

    2010-01-01

    This article explores the formation of citizenship as social practice in a school in El Alto, Bolivia. I examine interactions between "banking" forms of education, students' responses, and embodied practices of belonging and political agency, and argue that the seemingly passive forms of knowledge transmission so criticized by critical…

  12. Set of rules SOR 2 reactor site criteria

    International Nuclear Information System (INIS)

    1976-06-01

    The purpose of this set of rules is to describe criteria which guide the Director in his evaluation of the suitability of proposed sites for stationary power and testing reactors subject to SOR 2. (B.G.)

  13. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-01-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

  14. Litho-structural and geophysics features of the Alto Paranaiba Uplift

    International Nuclear Information System (INIS)

    Hasui, Y.

    1991-01-01

    The Alto Paranaiba Uplift (APU) is an almost elliptical tectonic feature of the Western Minas Gerais/Southern Goias region, which was active mostly during the Cretaceous. It separated the Parana Basin, during the formation of the Sao Bento, Uberaba and Bauru sequences, from the Alto-Sanfranciscana Basin, at the time of formation of the Areado, Patos, Capacete and Urucuia sequences. The Bouguer anomaly data indicate that the APU developed at the southwestern border of the ancient Brasilia crustal block and is represented by an almost elliptical gravity high of 15 mgal, locally disturbed by positive and negative the presence of important lineaments of a NW-SE set, mostly crossing the southwestern half of the APU. The APU development, the magmatism and the lateral basin formation involved reactivation of preexisting discontinuities and are related to a mantle plume. The tectonic development was aborted at the uplift stage during Cretaceous, after the deposition of the Bauru and Urucuia sequences, as is indicated by the Pratinha peneplane, now elevated at about 1.100 m altitude, which sculpture ended at the beginning of the Tertiary. The APU is one tectonic feature like other similar anomalies also aborted in the uplift stage or in the rift stage, which developed in Southern Brazil during the time of Atlantic Ocean opening. (author)

  15. Panstrongylus megistus em ecótopos artificiais de ilhas do Alto Rio Paraná

    Directory of Open Access Journals (Sweden)

    Guilherme Ana Lucia Falavigna

    2001-01-01

    Full Text Available Em resposta a denúncias de triatomíneos em ilhas do Alto Rio Paraná foram investigados 145 ecótopos artificiais e 4 (2,8% deles encontravam-se infestados: residência, "clube", ex-escola e monte de madeira. Foram analisados 17 de 35 P. megistus coletados; 12 (70,6% apresentavam-se infectados por Trypanosoma cruzi. Ave e roedor constituíram as fontes alimentares mais comuns. Todos os exames sorológicos (56 de humanos, 18 de cães e 10 de gatos foram negativos.

  16. Eventos adversos relacionados à terapia ventilatória em recém-nascidos de alto risco

    OpenAIRE

    França, Débora Feitosa de

    2016-01-01

    Objetivou-se analisar os eventos adversos relacionados à terapia respiratória em recém-nascidos de alto risco de uma unidade neonatal. Trata-se de um estudo observacional, longitudinal e prospectivo, realizado em uma maternidade, unidade de referencia no Estado do Rio Grande do Norte para gravidez e nascimento de alto risco. Os dados foram coletados no período de abril a setembro 2016, após aprovação do projeto no Comitê de Ética em Pesquisa da UFRN com CAAE nº 51832415.0.0000.5537. A amostra...

  17. Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

    International Nuclear Information System (INIS)

    Richard P. Wells

    2007-01-01

    This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

  18. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-11-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  19. Producción de miel e infestación con Varroa destructor de abejas africanizadas (Apis mellifera con alto y bajo comportamiento higiénico

    Directory of Open Access Journals (Sweden)

    Carlos Aurelio Medina-Flores

    2014-01-01

    Full Text Available El objetivo del estudio fue comparar la producción de miel y los niveles de Varroa destructor entre colonias de abejas africanizadas (AA ( Apis mellifera con alto y bajo comportamiento higiénico (CH en el altiplano semiárido de México. Se midió el nivel de CH a 57 colonias por congelamiento de la cría con nitrógeno líquido (N 2 . Las colonias se clasificaron en dos grupos: alto CH (> 95 % y bajo CH (0.05. Estos resultados sugieren que aparentemente el comportamiento higiénico no tiene un efecto mayor en la resistencia de las AA al crecimiento poblaci onal del ácaro. También sugieren que el comportamiento higiénico alto podría contribuir a incrementar la producción de miel en épocas del año con flujo reducido de néctar.

  20. CRECIMIENTO DEL MAÍZ EN VERTISOLES CON ALTO ALUMINIO EN LA BAIXADA MARANHENSE PRE-AMAZONIA, BRASIL

    Directory of Open Access Journals (Sweden)

    Alessandro Costa da Silva

    2014-01-01

    Full Text Available Crecimiento del maíz en vertisoles con alto aluminio en la Baixada Maranhense pre-Amazonia, Brasil. El objetivo de este trabajo fue evaluar el crecimiento del maíz en suelos con alto contenido de aluminio. Se midió el efecto del Al3+ en raíces y la cantidad de materia seca (raíz, hoja y tallo de maíz. Se efectuó la caracterización físico-química de cuatro muestras de suelo con alto aluminio colectadas del horizonte Ap, en tres municipios de la región conocida como Baixada Maranhense (Pre-Amazonia, Brasil: Santa Rita (SR, Arari (AR y Vitoria do Mearim (VM y un testigo colectado en el municipio de São Luís, Área del Núcleo de Tecnología Rural (T. El estudio, ejecutado en 2009, se llevó a cabo en invernadero y se utilizó 2 dm3 de suelo por maceta. Asimismo las muestras fueron divididas en muestras con y sin fertilización. La variación en la longitud de la raíz y de materia seca de las hojas difirió significativamente entre tratados con y sin fertilizante, excepto en la muestra de la localidad T. La producción de materia seca de raíz, tallo y hoja fue mayor en todos los suelos cuando se fertilizó. El suelo testigo también superó a todos los demás en cuanto a producción de materia seca en la raíz, posiblemente como resultado de una menor cantidad de Al3+ (1,2 cmolc/dm3 en comparación con los suelos SR, AR y VM (6,8; 8,0 y 7,0 cmolc/dm3 respectivamente. Se concluye la fertilización reduce el efecto detrimental del aluminio en la producción de maíz en la Baixada Maranhense.

  1. An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

    International Nuclear Information System (INIS)

    Harnden, A.M.C.

    1995-09-01

    The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

  2. Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

    International Nuclear Information System (INIS)

    Abou Yehia, H.; Bars, G.; Tran Dai

    2001-01-01

    After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

  3. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  4. Coal exploration in the Alto San Jorge area, Cordoba Department. Exploracion de carbones en el Ato San Jorge, Departamento de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Ospina, L H; Oquendo, G G [Geominas Ltda, Medellin (Colombia)

    1989-01-01

    A Mining Feasibility Study in the Area of Alto San Jorge, Department of Cordoba, Colombia, was commissioned by CARBOCOL S.A. to the Consortium Geominas-NACI. An area of 800 Ka2 was explored to define surface mining possibilities within two subareas referred to as Alto San Jorge and San Pedro Ure. Rocks of Cretaceous, Tertiary and Quaternary age crop out in the zone. In the subarea Alto San Jorge the principal structure is a syncline with a south-north direction. The San Pedro Ure subarea is formed by undulations with flanks of low dip, the most important being the San Antonio Syncline because it contains the mining block. The geological study of the surface demonstrated the existence of coal in the Oligocene Cienaga de Oro Formation and the Niocene Cerrito Formation, with potential resources of 6.3 billion tons. The subsequent exploration of the subsoil, with 20.618 m of drilling, permitted determination of demonstrated reserves in the order of 2.9 billion tons within two areas. In the sector selected for the mine plan, in the area of San Pedro-Puerto Libertador, 7.791 m of drilling was accomplished to define a demonstrated reserve of 515 million tons of coal down to a depth of 200. The combustible type coal has 5.000 cal/g. Complete mining schedules were developed at the prefeasibility level for two surface mines with productions of 1.5 MMTY and 4 MMTY. 9 figs., 3 tabs., 28 refs.

  5. Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

    1989-01-01

    A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

  6. 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

  7. Rapid data acquisition from the safety system of the FRJ-2 reactor

    International Nuclear Information System (INIS)

    Inhoven, H.

    1980-06-01

    The central department for research reactors (ZFR) of the Juelich Nuclear Research Centre (KFA) is operating the reactors FRJ-1 (MERLIN) and FRJ-2 (DIDO) since 1962. In 1976, a Siemens 330 computer has been put into operation especially for the processing of data from the DIDO reactor, followed by another computer of the same type for the purpose of processing data from the ZFR department in general. The present report is a result of the work investigating 'Data acquisition and data processing in the FRJ-2' and primarily discusses the complex of 'fast analog and binary signals'. The activities in this field of work have been and still are mainly concerned with general problems encountered in adapting a currently 14-year-old reactor system to a digital computer, namely problems such as data decoupling in the safety system of the reactor, data acquisition using the CAMAC system, data transfer via an 'extended branch', data acquisition software as core-resident programs, temporary storage as common data, interpreting software as peripheral - storage - resident programs. (orig./WB) [de

  8. Amplificador de Potencia de Alto Rendimiento para Transmisores EER

    OpenAIRE

    Ortega González, Francisco Javier; Gimeno Martín, Alejandro; Pardo Martin, José Manuel; Benavente Peces, César

    2008-01-01

    Se presenta un amplificador de potencia de alto rendimiento específicamente diseñado para aplicaciones EER (Envelope Elimination Restoration) en transmisores de HF. El amplificador se compone de dos subsistemas: Un amplificador clase-E de banda ancha para HF (B = 40%, POUT = 50W @ 7.5 MHz, ηOV > 90%) excitado por un driver también de banda ancha que amplifica la componente de fase de la señal y un amplificador de envolvente derivado de un amplificador clase-D de audio (o clase-S) que presenta...

  9. Rb-Sr measurements on metamorphic rocks from the Barro Alto Complex, Goias, Brazil

    International Nuclear Information System (INIS)

    Fuck, R.A.; Neves, B.B.B.; Cordani, U.G.; Kawashita, K.

    1988-01-01

    The Barro Alto Complex comprises a highly deformed and metamorphosed association of plutonic, volcanic, and sedimentary rocks exposed in a 150 x 25 Km boomerang-like strip in Central Goias, Brazil. It is the southernmost tip of an extensive yet discontinuous belt of granulite and amphibolite facies metamorphic rocks which include the Niquelandia and Cana Brava complexes to the north. Two rock associations are distinguished within the granulite belt. The first one comprises a sequence of fine-grained mafic granulite, hypersthene-quartz-feldspar granulite, garnet quartzite, sillimanite-garnet-cordierite gneiss, calc-silicate rock, and magnetite-rich iron formation. The second association comprises medium-to coarse-grained mafic rocks. The medium-grade rocks of the western/northern portion (Barro Alto Complex) comprise both layered mafic rocks and a volcanic-sedimentary sequence, deformed and metamorphosed under amphibolite facies conditions. The fine-grained amphibolite form the basal part of the Juscelandia meta volcanic-sedimentary sequence. A geochronologic investigation by the Rb-Sr method has been carried out mainly on felsic rocks from the granulite belt and gneisses of the Juscelandia sequence. The analytical results for the Juscelandia sequence are presented. Isotope results for rocks from different outcrops along the gneiss layer near Juscelandia are also presented. In conclusion, Rb-Sr isotope measurements suggest that the Barro Alto rocks have undergone at least one important metamorphic event during Middle Proterozoic times, around 1300 Ma ago. During that event volcanic and sedimentary rocks of the Juscelandia sequence, as well as the underlying gabbro-anorthosite layered complex, underwent deformation and recrystallization under amphibolite facies conditions. (author)

  10. 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.

  11. 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

  12. Estimation of power feedback parameters of pulse reactor IBR-2M on transients

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Popov, A.K.

    2013-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

  13. Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

    International Nuclear Information System (INIS)

    Stojic, M.; Pavicevic, M.

    1964-01-01

    This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

  14. PSA Level 2 activities for RBMK reactors

    International Nuclear Information System (INIS)

    Gubler, R.

    1998-01-01

    Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

  15. Studsvik's R2 reactor - Review of activities

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

    1993-07-01

    A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

  16. Signals in water - the deep originated CO2 in the Peschiera-Capone acqueduct in relation to monitoring of seismic activity in central Italy

    Directory of Open Access Journals (Sweden)

    Claudio Martini

    2017-01-01

    Full Text Available Valuation of the analysis performed on groundwater of Central Lazio by ACEA ATO2 SpA from 2001 to 2016, according to the model proposed by Chiodini et al. in 2004 that identifies in the Tyrrhenian coast of central and southern Italy, two notable releasing areas of the CO2 produced by the sub-crustal magma activity, or two areas of natural degassing of the planet: the TRDS area (Tuscan Roman degassing structure and the CDS area (Campanian degassing structure. Reconstruction of the CO2 produced by degassing through the analysis of the components of inorganic carbon measured in groundwater of Central Lazio (Rome and Rieti districts between 2001 and 2016. Causal relationship of the activity of mantle degassing in the TRDS area with the disastrous earthquake occurred at L’Aquila in April 6, 2009. Current use of the dissolved inorganic carbon measurement in the Peschiera and Capore spring waters to monitor the activity of mantle degassing in the TRDS area, in order to have an early warning signal of possible seismic activity in the Central Apennines. Revision and data updating after the earthquake in August 24, 2016 at Amatrice.

  17. El deslizamiento de Palo Alto, Turrialba, Costa Rica : apuntes para su estudio

    Directory of Open Access Journals (Sweden)

    Peraldo Huertas, Giovanni

    2015-12-01

    álisis de las morfologías del terreno se determinó que, dentro la forma típica en herradura que muestra este deslizamiento, se pueden distinguir 4 unidades: 1. Escarpes internos, 2. Bloques basculados, 3. Coronas laterales del deslizamiento y 4. Áreas de depósito de los materiales deslizados. El deslizamiento posee una serie de escarpes alargados, algunos llegan a medir aproximadamente 10 a 15 metros de altura. Al noroeste de Peralta se observan una serie de bloques, algunos alargados y otros con una forma más ovalada, los cuales se encuentran basculados en dirección NW, así como algunas zonas planas que, mediante comprobación de campo se determinó que corresponden en su mayoría con zonas anegadas. Se observan también colinas redondeadas, algunas alargadas, aisladas y en medio de ellas áreas deprimidas. Las morfologías de depositación presentan pendientes suaves, no mayores a los 10°, y onduladas. Entre los efectos que se generan por el movimiento en bloques en el área del deslizamiento, se encuentran repeticiones en la secuencia litológica y alteraciones en el patrón de drenaje existente. El hecho de que el río Guaitil discurra dentro del área del deslizamiento Palo Alto, hace un llamado de atención, ya que se observó la presencia de fisuras abiertas en su cauce que permiten la filtración de agua, lo que lleva a suponer que el deslizamiento esté en proceso de reactivación. Además, ante la posible reactivación económica que el proyecto Reventazón causará en el área, es importante generar más investigación para efectos de ordenamiento territorial y de gestión del riesgo This paper seeks to characterize Palo Alto landslide, from a geological and geomorphological point of view, in the context of mega-slope instability processes present along the Reventazón River. The Reventazón River corridor, between Turrialba and Siquirres, shows a series of complex mass movement processes, which generate continuous landslide related morphologies, giving

  18. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi

    2014-01-01

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  19. Reproduction of the PSBR reactor with Exterminator-2; Reproduccion del reactor PSBR con exterminador-2

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-08-15

    To reproduce the reactor PSBR reported in (1), with the available version of the Exterminator-II in the ININ, they took the dimensions, composition specifications, effective sections of the different compositions (excepting those of the central thimble and of the moderator), the K{sub eff} and the factors of power (FP) for the different burners. Based on the comparison of the K{sub eff} and of the FP obtained with those reported the precision it is determined before in the reproduction of the reactor mentioned. (Author)

  20. Calidad del coque de Horno Alto en la Unión Europea

    Directory of Open Access Journals (Sweden)

    Alvarez, R.

    2002-10-01

    Full Text Available After a brief review of the coking technology at the beginning of the new millennium, blast furnace coke quality criteria of most of EU countries, presented by the European Blast Furnace Committee in the 4th European Coke and Ironmaking Congress, are compared with those used by the Spanish Steel Industry at Aceralia. Blast furnace coke quality is very high in EU's countries in order to meet the requirements of bigger blast furnaces commissioned in the last years. CSR index is the most important parameter in the control of coke quality in Europe.

    En el presente trabajo se ha llevado a cabo una breve revisión de las tecnologías de coquización existentes al comienzo del nuevo milenio. Los criterios de calidad del coque de Horno Alto de la mayoría de los países de la Unión Europea, recogidos por el European Blast Furnace Committee, y que fueron presentados en el 4th European Coke and Ironmaking Congress en París durante el año 2000, se comparan con los utilizados por la industria siderúrgica española Aceralia. Como consecuencia del sensible aumento experimentado en el tamaño de los modernos Hornos Altos durante los últimos años, se ha podido comprobar que, en la UE, los valores de los diversos parámetros de control de calidad del coque son bastante similares y con unos requerimientos muy elevados. Asimismo, en la UE el parámetro CSR se ha convertido en el más importante para el control de la calidad del coque de Horno Alto.

  1. Production of Sn-117m in the BR2 high-flux reactor.

    Science.gov (United States)

    Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

    2009-01-01

    The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

  2. Evaluación de la exposición a selenio en Los Altos de Jalisco, México Evaluation of the exposure to selenium in Los Altos de Jalisco, México

    Directory of Open Access Journals (Sweden)

    Roberto Hurtado-Jiménez

    2007-08-01

    Full Text Available OBJETIVO: Evaluar la exposición a selenio (Se vía agua potable en los habitantes de Los Altos de Jalisco. MATERIAL Y MÉTODOS: Se determinó la concentración de Se en 125 pozos y se estimaron los niveles de exposición a Se en bebés, niños y adultos. RESULTADOS: La dosis de exposición y la ingestión de Se vía agua potable variaron en los siguientes rangos: a bebés: 1.3-6.7 µg/kg/d y 12.6-67.2 µg/d; b niños: 0.8-4.5 µg/kg/d y 16.8-89.6 µg/d; c adultos: 0.6-3.0 µg/kg/d y 33.6-179.2 µg/d. CONCLUSIONES: En este caso, la exposición a Se representa un riesgo potencial para la salud de la población, ya que en la mayoría de los casos es mayor que la recomendada por organismos internacionales de salud. Sin embargo, no es tan alta como para esperar la ocurrencia de selenosis.OBJECTIVE: To evaluate the exposure to selenium in drinking water in Los Altos de Jalisco (Jalisco State Heights. MATERIALS AND METHODS: The concentration of selenium was determined in 125 water wells, and the exposure doses to selenium were estimated for babies, children and adults. RESULTS: The estimated values of the exposure doses to selenium and total intake of selenium were in the following ranges, respectively: (a babies: 1.3-6.7 µg/kg/d and 12.6-67.2 µg/d; (b children: 0.8-4.5 µg/kg/d and 16.8-89.6 µg/d, (c adults: 0.6-3.0 µg/kg/d and 33.6-179.2 µg/d. CONCLUSIONS: The estimated exposure levels to selenium were higher than those recommended as optimum by international health organizations, representing a potential health risk. Nevertheless, estimated values are not high enough to produce selenosis.

  3. 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

  4. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  5. Comparative study between fluidized bed and fixed bed reactors in methane reforming with CO2 and O2 to produce syngas

    International Nuclear Information System (INIS)

    Jing Qiangshan; Lou Hui; Mo Liuye; Zheng Xiaoming

    2006-01-01

    Reforming of methane with carbon dioxide and oxygen was investigated over Ni/MgO-SiO 2 catalysts using fixed bed and fluidized bed reactors. The conversions of CH 4 and CO 2 in a fluidized bed reactor were close to thermodynamic equilibrium. The activity and stability of the catalyst in the fixed bed reactor were lower than that in the fluidized bed reactor due to carbon deposition and nickel sintering. TGA and TEM techniques were used to characterize the spent catalysts. The results showed that a lot of whisker carbon was found on the catalyst in the rear of the fixed bed reactor, and no deposited carbon was observed on the catalysts in the fluidized bed reactor after reaction. It is suggested that this phenomenon is related to a permanent circulation of catalyst particles between the oxygen rich and oxygen free zones. That is, fluidization of the catalysts in the fluidized bed reactor favors inhibiting deposited carbon and thermal uniformity in the reactor

  6. Estudio farmacoepidemiológico de uso de antiinflamatorios no esteroideos en pacientes de alto riesgo cardiovascular

    Directory of Open Access Journals (Sweden)

    Jorge Enrique Machado-Alba

    Full Text Available Con el objetivo de determinar la frecuencia de uso prolongado de antiinflamatorios no esteroideos (AINES en pacientes colombianos de alto riesgo cardiovascular (ARC se desarrolló un estudio retrospectivo en el cual se identificaron pacientes de ARC que usaron AINES por más de cinco meses continuos entre enero de 2011 y marzo de 2013. Se identificó a los pacientes que recibían crónicamente nitratos, digitálicos, y clopidogrel y ácido acetil salicílico (ASA, quienes fueron identificados como de ARC. Se realizó un análisis de frecuencias de uso según la comedicación recibida. Se encontró uso concomitante de AINES en el 0,35% de los consumidores de nitratos (tiempo promedio: 9,5 ± 4,4 meses, en el 0,36% de los consumidores de clopidogrel y ASA (tiempo promedio: 9,3 ± 3,4 meses, y en el 0,4% de los consumidores de digitálicos (10,2 ± 4,6 meses. Se concluye que existe una baja proporción de uso de AINES de manera crónica en pacientes de alto riesgo cardiovascular.

  7. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  8. Organismo territoriale e annodamenti urbani. Metodi di progetto per i centri minori del Lazio / Territorial organism and urban knottingt. Design methods for minor centers of Lazio

    Directory of Open Access Journals (Sweden)

    Giuseppe Strappa

    2013-07-01

    , con esisti disastrosi, a quello abitativo. / The research, currently being finalized by the study group by me coordinated at the Faculty of Architecture of "Sapienza" in Rome, within the more general topic of "City in extension", investigates the problem of how minor historical centers may experience contemporary transformations "congruent" with their formative process, in the belief that it is necessary to accept the incontrovertible fact that an urban organism, like any living organism, can only host continuous modifications. The study is part of a general framework of the national research and shares the assumption that the connotations of the Italian landscape suggests alternative tools, compared to the current ones, for the architectural design, and the possibility of an original placement, with specific characters, in the international state art of the discipline. In particular, the research examines the landscape characters of the Lazio region due to the diffusion in the territory of towns of considerable historical interest that are rapidly losing their quality. These centers are structurally weakened also by the influence of the near metropolitan area of Rome, wich its rapidly changing its shape through a gradual disorganization. The research proposes (and verifies the potential in some case studies, the reading of the urban fabrics of small towns, their form with typical characters, their potential “knotting”, the transformations in nodal places to form new nodal organisms, specialized building that innovate the existing fabric in a coherent and proportionate manner, allowing to avoid a “specialistic sprawl” (see the case of the displacement of the town halls outside of the town center that is added, with a disastrous result, to the current residential sprawl.

  9. Oxygen suppression in boiling water reactors. Phase 2. Annual report 1981, December 2, 1980-December 31, 1981

    International Nuclear Information System (INIS)

    Burley, E.L.

    1982-07-01

    A hydrogen addition test will be performed in the Dresden-2 reactor of Commonwealth Edison Company during 1982. Up to 2 ppM hydrogen will be added to and dissolved in the reactor feedwater to reverse the radiolysis reaction in the reactor core and suppress oxgen concentration in the primary coolant. At low oxygen levels the propensity of stressed and sensitized 304 stainless steel toward intergranular stress corrosion cracking is greatly reduced. The test will answer outstanding questions and uncertainties in the areas of water chemistry, equipment design and materials performance. Nine special sample facilities will be prepared in the primary coolant, main stream, feedwater/condensate, and offgas systems. Instrumentation will be available to measure hydrogen, oxygen, conductivity, pH, soluble and insoluble corrosion products, and electrochemical potentials. In addition, an autoclave in which confirming constant extension rate tests can be conducted in reactor water will be provided

  10. Characterization of fuel distributions in the Three-Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-04-01

    The resolution of technical issues generated by the accident at Three-Mile Island Unit 2 (TMI-2) will inevitably be of long range benefit. Determination of the fuel debris dispersal in the TMI-2 reactor system represents a major technical issue. In reactor recovery operations, such as for the safe handling and final disposal of TMI-2 waste, quantitative fuel assessments are being conducted throughout the reactor core and primary coolant system

  11. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    International Nuclear Information System (INIS)

    Bissani, M; O'Kelly, D S

    2006-01-01

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  12. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bissani, M; O' Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

  13. analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

    International Nuclear Information System (INIS)

    Elnokity, O.E.M.

    2006-01-01

    this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

  14. Event review: International Knapping Workshop, with Bruce Bradley, Fazenda Monte Alto, Dourado, SP (Brazil

    Directory of Open Access Journals (Sweden)

    Elisa Theodora Adriana van Veldhuizen

    2016-07-01

    Full Text Available The event took place from 3 till 8 July 2016 at Fazenda Monte Alto, Dourado, SP, Brazil. The aim of the course was to provide intensive knapping training in order to enhance analytical methods and procedures. This training was not only for students, but also professionals who were interested in the course. The course was given by Bruce Bradley (University of Exeter, who has extensive experience with Stone Age technologies and experimental archaeology. Mercedes Okumura (PPGArq, National Museum, Federal University of Rio de Janeiro and Astolfo G. M. Araujo (Museum of Archaeology and Ethnology, University of São Paulo organized the course, which was sponsored by Fazenda Monte Alto, Café Helena, and the British Academy, Newton Mobility Grants Scheme (NG140077. The workshop had 15 participants from Brazil, Uruguay, the Netherlands and Canada.

  15. General outline of the operation and utilization of the BR2 reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

    1978-01-01

    The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

  16. Benchmark testing of CENDL-2 for U-fuel thermal reactors

    International Nuclear Information System (INIS)

    Zhang Baocheng; Liu Guisheng; Liu Ping

    1995-01-01

    Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

  17. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

  18. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  19. Digital, remote control system for a 2-MW research reactor

    International Nuclear Information System (INIS)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

  20. Technical and scientific report of the Alto project

    International Nuclear Information System (INIS)

    Essabaa, S.; Gardes, D.; Grialou, D.; Ibrahim, F.; Le Scornet, J.C.

    2002-01-01

    The Alto project means the installation of an electron linear accelerator inside the experimental area of the tandem accelerator of the nuclear physics institute of Orsay (IPNO, France). This linear accelerator comes from CERN where it was operating as a pre-injector for LEP. This equipment will allow IPNO'teams to perform fast kinetics studies in a domain different from that of ELYSE accelerator. The time resolution will not be as high as that of ELYSE (picosecond) but will be sufficient (microsecond) to produce free radicals in aqueous and gaseous media. The main expectations of this installation can be classified according 3 axis: 1) basic research (mainly the study of nuclear matter through photo-fission, 2) research and development of accelerators (by providing a test bench for new high frequency systems and superconducting components), and 3) applied research for industry concerning: biochemistry under irradiation, radiation sensibility, DNA breaking, food and drug sterilization and behaviour of electronic components under irradiation. This rapport details the research program that could be achieved with this equipment, describes its contributions in terms of economic development, cooperation with industry, student training, and specifies the needed investment and the operating and maintenance costs. (A.C.)

  1. A nodal Grean's function method of reactor core fuel management code, NGCFM2D

    International Nuclear Information System (INIS)

    Li Dongsheng; Yao Dong.

    1987-01-01

    This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

  2. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1978-01-01

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  3. Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

    International Nuclear Information System (INIS)

    Guelol Oezdere, Oya

    2003-01-01

    Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

  4. Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

    International Nuclear Information System (INIS)

    Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

    2016-01-01

    Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

  5. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  6. Conservation and public presentation of the argaric site of Castellón Alto (Galera, Granada

    Directory of Open Access Journals (Sweden)

    Rodríguez-Ariza, M. Oliva

    2000-12-01

    Full Text Available The magnificent preservation of the archaeological site of Castellón Alto permitted reconstruction of the urbanism of this settlement and the life of its inhabitants. In addition to the necessary conservation, two interventions have been carried out with the principal objective of facilitating access, visiting, and the understanding of the site by the majority of the public. The first intervention happened in 1989 and the main task was centered on the consolidation, restoration, and delimiting of the archaeological bed. The second one happened in 1997 and was centered in the consolidation and reconstruction of both a hut and two tombs. With the opening of the Archaeological Museum of Galera, the cultural and touristic contribution of Castellón Alto will be complete. It will provide an interpretation of this prehistoric village, as well as the Argaric culture in general and all the other archaeological sites of the area.

    La magnífica conservación del registro arqueológico del Castellón Alto permitía reconstruir el urbanismo del poblado y la vida de estas poblaciones. Se han efectuado dos actuaciones con el objetivo principal de facilitar, además de la necesaria conservación, el acceso, la visita y la comprensión del poblado prehistórico por parte de un público mayoritario. La primera actuación se realizó en 1989 y los trabajos se centraron principalmente en la consolidación, restauración y cerramiento del área del yacimiento. La segunda se realizó en 1997 y se centró en el acondicionamiento y reconstrucción de una cabaña y dos sepulturas. La oferta turística y cultural que ofrece el Castellón Alto se completará con la próxima apertura del Museo Arqueológico de Galera, donde se efectuará una interpretación de este poblado y de la cultura argárica, así como del resto de yacimientos de la zona.

  7. Relación entre el estilo de vida de una joven deportista de alto rendimiento y los patrones funcionales de salud de Marjory Gordon

    OpenAIRE

    Fabra Heredia, Juan Manuel; Casadó Marín, Lina

    2014-01-01

    Estudio de caso que busca conocer y comprender la relación que hay entre el estilo de vida de una joven deportista de alto rendimiento y los patrones funcionales de salud, a través de la valoración realizada a una joven deportista de alto rendimiento, desde una perspectiva holística, para adentrarse en las peculiaridades propias de este estilo de vida y crear un punto de partida para los cuidados de enfermería dirigidos a deportistas de alto rendimiento. En este caso, se observaron factores p...

  8. Maximum credible accident analysis for TR-2 reactor conceptual design

    International Nuclear Information System (INIS)

    Manopulo, E.

    1981-01-01

    A new reactor, TR-2, of 5 MW, designed in cooperation with CEN/GRENOBLE is under construction in the open pool of TR-1 reactor of 1 MW set up by AMF atomics at the Cekmece Nuclear Research and Training Center. In this report the fission product inventory and doses released after the maximum credible accident have been studied. The diffusion of the gaseous fission products to the environment and the potential radiation risks to the population have been evaluated

  9. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    Energy Technology Data Exchange (ETDEWEB)

    Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

    2016-03-11

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

  10. Numerical modelling of Alto Verde landslide using the material point method

    Directory of Open Access Journals (Sweden)

    Marcelo Alejandro Llano-Serna

    2015-01-01

    Full Text Available Finalizando el año 2008 en la ciudad de Medellín, Colombia, ocurrió un deslizamiento de tierra en la urbanización Alto Verde provocando la muerte de doce personas y la destrucción de seis viviendas. Los deslizamientos se destacan por el elevado nivel de deformaciones en una masa de suelo. El presente trabajo utilizó el método del punto material (MPM, método basado en partículas que utiliza una doble discretización Lagrangiano-Euleriana. La doble discretización genera un marco numérico robusto que permite la simulación de grandes distorsiones. El modelo numérico planteó una simplificación de las condiciones geotécnicas, morfológicas y estructurales de las edificaciones envueltas en Alto Verde. El estado de deformación final de la simulación se acomodó satisfactoriamente a las características geométricas finales observadas en campo. Los resultados obtenidos generan aplicaciones como el diseño de barreras, análisis de riesgo o la determinación de la distancia mínima de retiro a una ladera susceptible de deslizamiento.

  11. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  12. HTLV-I en población de alto riesgo sexual de Pisco, Ica, Perú.

    Directory of Open Access Journals (Sweden)

    Patricia GARRIDO

    1997-07-01

    Full Text Available Objetivo: Se estudiaron 141 personas con alto riesgo sexual en la ciudad de Pisco para detectar infección por HTLV-I. Material y Métodos: Se encuestaron y se tomaron muestras de sangre a 141 personas que involucró a trabajadoras sexuales (32, varones homosexuales (54, y varones bisexuales(55. Resultados: Tres de treintidós (10.4% trabajadoras sexuales fueron positivas; uno de cincuenticuatro (1.9% de varones homosexuales y ninguno de 55 bisexuales. Hubo una elevada frecuencia de parejas, así como el antecedente de enfermedades de transmisión sexual (ETS en estos grupos con comportamiento de riesgo. Conclusiones: El HTLV-I es una infección frecuente en grupos de alto riesgo sexual de Pisco-Perú. (Rev Med Hered 1997; 8:104-107.

  13. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  14. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  15. 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.)

  16. A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

    2014-01-01

    A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

  17. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  18. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  19. Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

    2010-01-01

    In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

  20. Avaliação da conduta conservadora na lesão intraepitelial cervical de alto grau

    Directory of Open Access Journals (Sweden)

    Nelson Shozo Uchimura

    2012-06-01

    Full Text Available OBJETIVO: Analisar a associação entre a conduta conservadora em lesão intraepitelial cervical de alto grau com o índice de recidiva da neoplasia e faixa etária. MÉTODOS: Estudo transversal e retrospectivo realizado com 509 mulheres (15-76 anos atendidas no período de 1996 a 2006, com colpocitologia oncótica alterada, em um serviço público de referência em Maringá, PR. Os dados foram coletados dos prontuários médicos e estudadas as variáveis diagnóstico definitivo, tipos de tratamento, ocorrência da lesão e recidivas, analisados por meio de testes de associação de qui-quadrado de Pearson e teste exato de Fisher. RESULTADOS: A lesão intraepitelial cervical de alto grau ocorreu em 168 casos; destes, 31 mulheres foram submetidas à amputação cônica, 104 a cirurgias de alta frequência, nove histerectomizadas e 24 receberam conduta conservadora. Dentre as mulheres com lesão de alto grau e tratadas de forma conservadora, oito (33,3% recidivaram, enquanto dentre as submetidas à conduta não conservadora dez (6,9% recidivaram, sendo essa diferença estatisticamente significante (p = 0,0009, RP = 4,8 (IC95% 2,11;10,93. Para aquelas que fizeram o seguimento clínico-citológico, três (30,0% e, dentre as cauterizadas, cinco (35,7% recidivaram no prazo de três anos, sem diferença significante (p = 0,5611. A recidiva abaixo e acima de 30 anos ocorreu, respectivamente, em sete (13,8% e 11 (12,2% mulheres (p = 0,9955. CONCLUSÕES: A idade da mulher não influencia o prognóstico de recidiva. O tratamento conservador deve ser indicado como conduta de exceção, dada a alta taxa de recidiva, e o seguimento deve ser rigoroso, com acompanhamento citológico e colposcópico de até três anos, período em que ocorre a maioria das recidivas.

  1. Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

    2013-01-01

    This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

  2. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

  3. Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

    International Nuclear Information System (INIS)

    Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

    1999-01-01

    Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

  4. An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

    Energy Technology Data Exchange (ETDEWEB)

    Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2017-01-15

    Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

  5. 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

  6. Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant.

  7. Floristic Diversity of Two Zones of Humid Tropical Forest at Alto Baudó, Chocó, Colombia

    Directory of Open Access Journals (Sweden)

    Luis Javier Mosquera Ramos

    2007-08-01

    Full Text Available Between June and August of 2005 the floristic composition ≥1 cm of DAP was determined in an area of ? 0.2 ha of humid tropical forest at the localities of Pie de >Pató (05º 30' 56" N and 76º 58' 26" W and Nauca (5º 41' 6" N and 77º 00' 36" W, Alto Baudó, Chocó Colombia . En each locality an area of 0.1 ha was sampled which was divided into smaller areas of 2 x 50 cm each. A total of 1618 inidivduals were recorded represented by 257 species, 156 genres and 56 botanical families from which 842 individuals, 161 species, 108 genres and 46 families where found at Pie de Pató, and 776 individuals, 161 species, 98 genres and 45 families at Nauca. At Pie de Pató the families best represented in terms of genres were Rubiaceae (12 genres and 27 species, Arecaceae (eight genres and eight species and Bombacaceae (seven genres and ten species. At Nauca they were Rubiaceae (eleven genres and 25 species, Moraceae (eight genera and 13 species and Arecaceae (eigth genres and eight species. The richness index was of 23,75 and 24,05 for Pie de Pató and Nauca respectively. Diversity change was stimated as 4,43 for both localities. These results indicate high diversity of these forests at Alto Baudó.

  8. Problems of nuclear reactor safety. Vol. 2

    International Nuclear Information System (INIS)

    Goncharov, L.A.

    1995-01-01

    Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

  9. The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

    International Nuclear Information System (INIS)

    Aziz, M.; El Messiry, A.M.

    2000-01-01

    The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

  10. Diversity and similarity of trophic system "Barn Owl - terrestrial mammals" in the volcanic districts of Latium (Italy / Diversità ed affinità dei sistemi trofici "Tyto alba - mammiferi terragnoli" nei comprensori vulcanici del Lazio

    Directory of Open Access Journals (Sweden)

    Fiorella Aste

    1987-07-01

    Full Text Available Abstract Bony remains of about ten thousands small terrestrial mammals preyed by Barn Owl in six volcanic districts of Latium were examined and relevant biocoenotic parameters (such as biotic diversity, thermoxerophily index, Renkonen's and Faith's indexes calculated. Diversity values exhibit no apparent correlation with a number of environmental and biocoenotic parameters of non-anthropic origin - i.e.: district age, height on sea level, latitude, biocoenotic (Renkonen's and faunistic (Faith's affinities. Conversely, a clearly significant, negative correlation with landscape anthropization was shown, revealing the importance of man's impact in shaping functional connections in the terrestrial communities of studied region. Riassunto L'esame del sistema trofico in argomento in 6 distretti vulcanici del Lazio ha posto in evidenza che la diversità biotica è significativamente e inversamente correlata con l'antropizzazione territoriale, ma non con altri fattori ambientali di origine anantropica.

  11. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

  12. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    International Nuclear Information System (INIS)

    Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

    2012-01-01

    CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

  13. Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

    Science.gov (United States)

    Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

    2012-08-01

    CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

  14. Utilization of the SLOWPOKE-2 research reactor

    International Nuclear Information System (INIS)

    Lalor, G.C.

    2001-01-01

    SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

  15. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  16. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

    International Nuclear Information System (INIS)

    Markovic, H.

    1962-07-01

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included [sr

  17. Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Souza Mendes, J.E. de

    1987-01-01

    Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

  18. CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

    International Nuclear Information System (INIS)

    Barelli, L.; Bidini, G.; Gallorini, F.

    2016-01-01

    Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

  19. MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

    Directory of Open Access Journals (Sweden)

    Abdul Wahid

    2015-06-01

    Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

  20. Operating reactors licensing actions summary. Volume 5, No. 2

    International Nuclear Information System (INIS)

    1985-04-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

  1. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, R; Aalto, E

    1964-09-15

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

  2. Ventaja de jugar en casa en voleibol de alto rendimiento

    Directory of Open Access Journals (Sweden)

    Rui Marcelino

    2009-01-01

    Full Text Available En este estudio se ha pretendido estudiar la ventaja en casa en el Voleibol de alto rendimiento, apoyándonos en las estadísticas de los partidos que más pueden explicar ese fenómeno. Se han analizado 65.949 acciones de juego seleccionadas de la Liga Mundial 2005. El instrumento de observación que se ha elaborado es una combinación de formato de campo con sistemas de categorías. Los datos se han obtenido mediante el "Volleyball Information System" y se han analizado mediante la prueba t de Student, y la prueba ji-cuadrado. Los resultados demuestran que hay ventaja en casa en Voleibol (57,5% de victorias jugando en casa. Las estadísticas del ataque (t= 2.49, p = 0.01, del servicio (t= -2.18, p= 0.03, de la recepción (t= 16.74, p<0.001 y de la distribución (t= 2.03, p= 0.04 muestran rendimientos superiores para los equipos que juegan en casa. No se han encontrado diferencias en el rendimiento del bloqueo (t= -0.25, p= 0.80 y la defensa (t= 0.11, p= 0.92 entre los juegos disputados en casa y los disputados fuera.

  3. Microflow photochemistry: UVC-induced [2 + 2]-photoadditions to furanone in a microcapillary reactor

    Directory of Open Access Journals (Sweden)

    Sylvestre Bachollet

    2013-10-01

    Full Text Available [2 + 2]-Cycloadditions of cyclopentene and 2,3-dimethylbut-2-ene to furanone were investigated under continuous-flow conditions. Irradiations were conducted in a FEP-microcapillary module which was placed in a Rayonet chamber photoreactor equipped with low wattage UVC-lamps. Conversion rates and isolated yields were compared to analogue batch reactions in a quartz test tube. In all cases examined, the microcapillary reactor furnished faster conversions and improved product qualities.

  4. A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

    International Nuclear Information System (INIS)

    Abd El-Rahman, M.F.

    2004-01-01

    Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

  5. Thermal design of heat-exchangeable reactors using a dry-sorbent CO2 capture multi-step process

    International Nuclear Information System (INIS)

    Moon, Hokyu; Yoo, Hoanju; Seo, Hwimin; Park, Yong-Ki; Cho, Hyung Hee

    2015-01-01

    The present study proposes a multi-stage CO 2 capture process that incorporates heat-exchangeable fluidized-bed reactors. For continuous multi-stage heat exchange, three dry regenerable sorbents: K 2 CO 3 , MgO, and CaO, were used to create a three-stage temperature-dependent reaction chain for CO 2 capture, corresponding to low (50–150 °C), middle (350–650 °C), and high (750–900 °C) temperature stages, respectively. Heat from carbonation in the high and middle temperature stages was used for regeneration for the middle and low temperature stages. The feasibility of this process is depending on the heat-transfer performance of the heat-exchangeable fluidized bed reactors as the focus of this study. The three-stage CO 2 capture process for a 60 Nm 3 /h CO 2 flow rate required a reactor area of 0.129 and 0.130 m 2 for heat exchange between the mid-temperature carbonation and low-temperature regeneration stages and between the high-temperature carbonation and mid-temperature regeneration stages, respectively. The reactor diameter was selected to provide dense fluidization conditions for each bed with respect to the desired flow rate. The flow characteristics and energy balance of the reactors were confirmed using computational fluid dynamics and thermodynamic analysis, respectively. - Highlights: • CO 2 capture process is proposed using a multi-stage process. • Reactor design is conducted considering heat exchangeable scheme. • Reactor surface is designed by heat transfer characteristics of fluidized bed

  6. 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%

  7. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

    2002-01-01

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  8. 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

  9. 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.

  10. 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

  11. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    Ponsard, B.

    2005-01-01

    The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

  12. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  13. A regression approach for Zircaloy-2 in-reactor creep constitutive equations

    International Nuclear Information System (INIS)

    Yung Liu, Y.; Bement, A.L.

    1977-01-01

    In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

  14. Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

    International Nuclear Information System (INIS)

    Elliot, A.J.; Stuart, C.R.

    2008-06-01

    An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

  15. Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

    2014-01-01

    Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

  16. Perceção do estado de saúde da população idosa do Alto Minho

    OpenAIRE

    Fernandes, Fábia de Jesus Felgeuiras

    2015-01-01

    Dissertação de mestrado em Gestão das Organizações: Ramo de Gestão de Unidades de Saúde (parceria com a APNOR) apresentada na Escola Superior de Saúde do Instituto Politécnico de Viana do Castelo Este estudo teve como objetivo avaliar o estado de saúde da população idosa do Alto Minho de forma a contribuir para o planeamento em saúde e emerge do projeto “Estado de Saúde e Atividade Física da População Idosa do Alto Minho”, financiado por Fundos FEDER através do Programa Operacional Fatores...

  17. Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

    Science.gov (United States)

    Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

    2014-01-01

    The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

  18. Direct In Situ Quantification of HO2 from a Flow Reactor.

    Science.gov (United States)

    Brumfield, Brian; Sun, Wenting; Ju, Yiguang; Wysocki, Gerard

    2013-03-21

    The first direct in situ measurements of hydroperoxyl radical (HO2) at atmospheric pressure from the exit of a laminar flow reactor have been carried out using mid-infrared Faraday rotation spectroscopy. HO2 was generated by oxidation of dimethyl ether, a potential renewable biofuel with a simple molecular structure but rich low-temperature oxidation chemistry. On the basis of the results of nonlinear fitting of the experimental data to a theoretical spectroscopic model, the technique offers an estimated sensitivity of reactor exit temperature range of 398-673 K. Accurate in situ measurement of this species will aid in quantitative modeling of low-temperature and high-pressure combustion kinetics.

  19. 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)

  20. Status and perspective of development of cold moderators at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Kulikov, S; Shabalin, E

    2012-01-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  1. Status and perspective of development of cold moderators at the IBR-2 reactor

    Science.gov (United States)

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  2. 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.

  3. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  4. Näljastreiki pidanud Mussolini ei suutnud oma parteid päästa / Ülle Toode

    Index Scriptorium Estoniae

    Toode, Ülle, 1969-

    2005-01-01

    Benito Mussolini lapselaps Alessandra Mussolini korraldas Roomas kohtuhoone ees näljastreigi, lootes, et kohus lubab tal osaleda Lazio maakonna kohalikel valimistel. Itaalia valimiskomisjoni teatel on Lazios Mussolini partei toetuseks kogutud rohkem kui 4000 allkirjast vähemalt 870 võltsitud

  5. 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

  6. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  7. Further study on parameterization of reactor NAA: Pt. 2

    International Nuclear Information System (INIS)

    Tian Weizhi; Zhang Shuxin

    1989-01-01

    In the last paper, Ik 0 method was proposed for fission interference corrections. Another important kind of interferences in reator NAA is due to threshold reaction induced by reactor fast neutrons. In view of the increasing importance of this kind of interferences, and difficulties encountered in using the relative comparison method, a parameterized method has been introduced. Typical channels in heavy water reflector and No.2 horizontal channel of Heavy Water Research Reactor in the Insitute of Atomic Energy have been shown to have fast neutron energy distributions (E>4 MeV) close to primary fission neutron spectrum, by using multi-threshold detectors. On this basis, Ti foil is used as an 'instant fast neutron flux monitor' in parameterized corrections for threshold reaction interferences in the long irradiations. Constant values of φ f /φ s = 0.70 ± 0.02% have been obtained for No.2 rabbit channel. This value can be directly used for threshold reaction inference correction in the short irradiations

  8. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  9. 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

  10. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

    2017-01-01

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  11. Osteossarcomas humanos de alto grau: imunoexpressão de p53, erb-2 e P-glicoproteína, e correlação com o parâmetro anaplasia

    OpenAIRE

    Alves,Maria Teresa de Seixas; Miiji,Luciana Nakao Odashiro; Marinho,Larissa Cardoso; Petrilli,Antonio Sergio; Jesus-Garcia,Reynaldo; Tolledo,Silvia Caminada; Patricio,Francy Reis da Silva

    2008-01-01

    INTRODUÇÃO: Osteossarcoma (OS), o mais freqüente tumor primário maligno do osso, tem comportamento local agressivo e alto índice de disseminação sistêmica. Os eventos que permitem o crescimento e a disseminação tumoral ainda permanecem controversos. Os estudos sobre a carcinogênese e a progressão dessa neoplasia, com base na imunoexpressão de c-erb-B2, P-glicoproteína (P-gp) e p53, apresentam resultados conflitantes acerca do real valor prognóstico e suas correlações com parâmetros histológic...

  12. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  13. TiO2-photocatalyzed As(III) oxidation in a fixed-bed, flow-through reactor.

    Science.gov (United States)

    Ferguson, Megan A; Hering, Janet G

    2006-07-01

    Compliance with the U.S. drinking water standard for arsenic (As) of 10 microg L(-1) is required in January 2006. This will necessitate implementation of treatment technologies for As removal by thousands of water suppliers. Although a variety of such technologies is available, most require preoxidation of As(III) to As(V) for efficient performance. Previous batch studies with illuminated TiO2 slurries have demonstrated that TiO2-photocatalyzed AS(III) oxidation occurs rapidly. This study examined reaction efficiency in a flow-through, fixed-bed reactor that provides a better model for treatment in practice. Glass beads were coated with mixed P25/sol gel TiO2 and employed in an upflow reactor irradiated from above. The reactor residence time, influent As(III) concentration, number of TiO2 coatings on the beads, solution matrix, and light source were varied to characterize this reaction and determine its feasibility for water treatment. Repeated usage of the same beads in multiple experiments or extended use was found to affect effluent As(V) concentrations but not the steady-state effluent As(III) concentration, which suggests that As(III) oxidation at the TiO2 surface undergoes dynamic sorption equilibration. Catalyst poisoning was not observed either from As(V) or from competitively adsorbing anions, although the higher steady-state effluent As(III) concentrations in synthetic groundwater compared to 5 mM NaNO3 indicated that competitive sorbates in the matrix partially hinder the reaction. A reactive transport model with rate constants proportional to incident light at each bead layer fit the experimental data well despite simplifying assumptions. TiO2-photocatalyzed oxidation of As(III) was also effective under natural sunlight. Limitations to the efficiency of As(III) oxidation in the fixed-bed reactor were attributable to constraints of the reactor geometry, which could be overcome by improved design. The fixed-bed TiO2 reactor offers an environmentally

  14. Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

    International Nuclear Information System (INIS)

    Skogen, F.B.; Stout, R.B.

    1977-01-01

    Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

  15. Structure of neutron rich nuclei of Germanium and Gallium beyond N equals 50 at Alto

    International Nuclear Information System (INIS)

    Lebois, M.

    2008-09-01

    The gamma rays following the beta decay of the following very neutron-rich isotopes: 82,83,84 Ga produced by photo-fission, have been studied at the newly built ISOL facility in Orsay: ALTO. In ALTO the interaction of an electron beam with U 238 target generates a continuous spectra of Bremsstrahlung gamma radiation that triggers U 238 fission. The fission fragments are then ionized, extracted and mass-separated. The analysis of the data has shown the existence of an isomer in 31 84 Ga 53 and has enabled us to confirm known results on 32 83 Ge 51 energy levels including the gamma transition between the 1/2+ state at 247,7 KeV and the fundamental state. We have also proposed the first energy level scheme for 33 84 As 51 . In order to understand the structure of the nucleus we have used the Thankappan and True model that gives a description of the coupling between the pair-pair core (half-magical) and the single nucleon. This model applied to the N=51 chain ( 38 89 Sr 51 , 36 87 Kr 51 , 34 85 Se 51 , 32 83 Ge 51 and 30 81 Zn 51 ) has allowed us to see the main features of odd isotope structure. We have also confirmed previous results concerning the nature of the states in the following decay 31 83 Ga 52 → 32 83 Ge 51

  16. Characterization of fuel distribution in the Three Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

    1984-01-01

    Neutron and gamma-ray dosimetry are being used for nondestructive assessment of the fuel distribution throughout the Three Mile Island Unit 2 (TMI-2) reactor core region and primary cooling system. The fuel content of TMI-2 makeup and purification Demineralizer A has been quantified with Si(Li) continuous gamma-ray spectrometry and solid-state track recorder (SSTR) neutron dosimetry. For fuel distribution characterization in the core region, results from SSTR neutron dosimetry exposures in the TMI-2 reactor cavity are presented. These SSTR results are consistent with the presence of a significant amount of fuel debris, equivalent to several fuel assemblies or more, lying at the bottom of the reactor vessel. (Auth.)

  17. Estudio microbiológico de los alimentos elaborados en comedores colectivos de alto riesgo

    Directory of Open Access Journals (Sweden)

    Pérez-Silva García Mª del Carmen

    1998-01-01

    Full Text Available FUNDAMENTO: Valorar los resultados del análisis microbiológico de los alimentos preparados en comedores colectivos de alto riesgo, con el fin de conocer el grado de contaminación de los alimentos, analizar las causas de dicha contaminación y mejorar la situación sanitaria de estos establecimientos. MÉTODOS: Estudio observacional descriptivo con los datos obtenidos de la inspección sanitaria en 44 comedores colectivos de alto riesgo, que incluyó el análisis microbiológico de 90 alimentos, así como la inspección sanitaria de los establecimientos. RESULTADOS: En los colegios los microorganismos mesófilos fueron los contaminantes más frecuentes; en las guarderías y residencias de ancianos predominaron los indicadores de higiene deficiente en la manipulación de alimentos. Los microorganismos mesófilos se encontraron durante los meses fríos en mayor proporción que durante los meses cálidos. Los indicadores de higiene deficiente aparecieron generalmente en los alimentos preparados en establecimientos en los que se observaron deficiencias. Los microorganismos psicrótrofos no se encontraron en ninguno de los alimentos recogidos en guarderías y sí en colegios y residencias de ancianos. CONCLUSIONES: Este estudio indica qué problemas predominan en cada tipo de comedor colectivo de alto riesgo. Los mesófilos aparecen en los alimentos elaborados en cocinas de tamaño grande, los indicadores de higiene deficiente se encontraron asociados a una manipulación de alimentos por personal no profesional y a establecimientos con deficiencias, y los psicrótrofos se detectaron en aquellos establecimientos que guardan la comida sobrante. Se sugieren recomendaciones para la eliminación de los problemas detectados.

  18. Simulación clínica de alto realismo: una experiencia en el pregrado

    Directory of Open Access Journals (Sweden)

    Javier Riancho

    Full Text Available Introducción. La simulación con modelos de alto realismo se utiliza a menudo en la formación de los profesionales sanitarios. Sin embargo, son escasas las experiencias en el pregrado. El objetivo de este trabajo fue conocer la factibilidad y la aceptación de su aplicación con estudiantes de sexto curso de la licenciatura de Medicina. Materiales y métodos. Se diseñaron ocho escenarios que simulaban problemas clínicos frecuentes para su desarrollo con maniquíes de alto realismo. Los estudiantes se dividieron en grupos de 6-8 sujetos, cada uno de los cuales atendió dos casos durante 30 minutos. Posteriormente se llevó a cabo un análisis reflexivo durante 25-40 minutos. La actividad se repitió en dos años consecutivos. Al final se recabó la opinión de los estudiantes mediante encuestas anónimas. Resultados. La actividad fue valorada muy positivamente por los estudiantes, quienes la consideraron como "útil" (4,8 y 4,9 puntos sobre 5 e "interesante" (4,9 y 4,9 puntos. El tiempo preciso para preparar cada escenario fue de unas 3 horas. Fueron necesarias una jornada completa de un profesor, un técnico y un enfermero para que un colectivo de unos 40 estudiantes se expusiera a dos casos clínicos. Conclusiones. Esta experiencia piloto sugiere que la simulación de alto realismo es factible en el pregrado, supone un consumo razonable de recursos y tiene una elevada aceptación por parte de los estudiantes. No obstante, se necesitan otros estudios que confirmen la impresión subjetiva de que resulta útil para potenciar el aprendizaje de los alumnos y su competencia clínica.

  19. Estimation of power feedback parameters of the IBR-2M reactor by square wave reactivity

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.; Sumkhuu, D.

    2016-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) are estimated based on the analysis of power transients caused by deliberate square wave reactivity when the pulsed reactor operates in the self-regulation mode. The PFB of the IBR-2M is described by three linear first-order differential equations. Two components of the PFB are responsible for the negative feedback and one, for the positive. The overall feedback is negative, i.e., it has a stabilizing effect for the operation of the reactor. The slowest negative component of the PFB is probably caused by heating of the fuel. Periodically repeated in the process of exploitation, estimation of the PFB parameters is one of the methods to ensure safety operation of the reactor. [ru

  20. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

  1. TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Hatch, A.J.

    1975-01-01

    1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

  2. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  3. Fuel dynamics by using Landscape Ecology Indices in the Alto Mijares, Spain

    Science.gov (United States)

    Iqbal, J.; Garcia, C. V.

    2009-04-01

    Land abandonment in Mediterranean regions has brought about a number of management problems, being an increased wildfire activity prevalent among them. Agricultural neglect in highlands resulted in reduced anthropogenic disturbances and greater landscape homogeneity in areas such as the Alto Mijares in Spain. It is widely accepted that processes like forest fires, influence structure of the landscape and vice versa. Fire-prone Mediterranean flora is well adapted to this disturbance, exhibiting excellent succession capabilities; but higher fuel loads and homogeneous conditions may ally to promote vegetation recession when the fire regime is altered by land abandonment. Both succession and recession make changes to the landscape structure and configuration. However, these changes are difficult to quantify and characterize. If landscape restoration of these forests is a management objective, then developing a quantitative knowledge base for landscape fuel dynamics is a prerequisite. Four classified LandsatTM satellite images were compared to quantify changes in landscape structure between 1984 and 1998. An attempt is made to define landscape level dynamics for fuel development after reduced disturbance and fuel accumulation that leads to catastrophic fires by using landscape ecology indices. By doing so, indices that best describe the fuel dynamics are pointed. The results indicate that low-level disturbance increases heterogeneity, thus lowers fire hazard. No disturbance or severe disturbance increases homogeneity because of vegetation succession and may lead to devastating fires. These fires could be avoided by human induced disturbance like controlled burning, harvesting, mechanical works for fuel reduction and other silviculture measures; thus bringing in more heterogeneity in the region. The Alto Mijares landscape appears to be in an unstable equilibrium where succession and recession are at tug of war. The effects are evident in the general absence of the climax

  4. Report on the operation in 1973 of the FR 2 research reactor

    International Nuclear Information System (INIS)

    Moeller, I.; Steiger, W.

    1975-04-01

    Also in 1973, the heavy-water moderated research and testing reactor FR 2 was operated to schedule at 44 MW nominal power. Again, the availability of the plant was slightly improved. Experimental utilization through instrumented irradiation capsules strongly increased as compared to the previous year. Up to 16 capsule test rigs at a time were inserted in the reactor. As to the beam tube experiments, up to 13 experiments covering a total of 18 test rigs were conducted simultaneously at the 12 reasonably usable beam holes. At the beginning of the year all of the positions available were occupied by 5 loop experiments. Isotope production reached its highest value with a total of 2,372 irradiated capsules (1.3% more than the year before). Some remarkable figures characterized the year 1973: On August 16, 1973 ten years of full power operation at a nominal power of 12 and 44 MW, respectively, had been reached. On July 24, 1973 the 50,000th isotope irradiation was performed in the reactor and on December 26, 1973 a total energy release of 100,000 MWd was recorded. Moreover, the 125,000th visitor of the reactor was welcomed on December 6, 1973. (orig./UA) [de

  5. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    International Nuclear Information System (INIS)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

  6. Esquistossomose mansoni autóctone e outras parasitoses intestinais em escolares do bairro alto da Boa Vista, da Cidade do Rio de Janeiro

    Directory of Open Access Journals (Sweden)

    Virginia T. Schall

    1985-09-01

    Full Text Available Procedeu-se a um levantamento coproscópico em alunos de 8 a 15 anos, pertencentes a 3 escolas do Alto da Boa Vista. De 155 alunos examinados, encontraram-se 4 casos positivos (2,6% de esquistossomose. Entre 25 familiares destes, encontraram-se outros 4 casos positivos (16%. Havia na área uma população de Biomphalaria tenagophila (Orbigny, 1835, amplamente distribuída nas valas de irrigação de hortas de agrião, ligadas ao rio das Fumas. A taxa de infecção dos moluscos foi de 7 (0,29% em 2.400 examinados. Comparando-se estes resultados com dados anteriores, observa-se que a prevalência da esquistossomose se mantém no Alto da Boa Vista há pelo menos 15 anos.The present situation regarding schistosomiasis mansoni prevalence among primary school children was evaluated in the periphery of Rio de Janeiro city (Alto da Boa Vista. A coproscopic survey was carried out with students of 8 to 15 years oldfrom 3 schools of the area. Among 15 5 examined, 4 positive cases (2,6% of schistosomiasis were found. On examining 25 relatives of the positive students, 4 other positives cases (16% were found, all from the same region. The malacological survey carried out revealed a Biomphalaria tenagophila population largely distributed in the irrigation ditches of water-cress gardens surrounding the Fumas River. Seven (7 out of2400 snails collected were positive for Schistosoma mansoni (0,29%. Comparing these results with previous data, it can be observed that schistosomiasis prevalence has been maintained in this area for as long as 15 years. Stool examinations revealed a high prevalence of A. lumbricoides (48,4%. E. nana (23,5% and T. trichiura (38%.

  7. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    Energy Technology Data Exchange (ETDEWEB)

    Dougherty, D.; Adams, J. W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

  8. Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

    International Nuclear Information System (INIS)

    Dougherty, D.; Adams, J.W.

    1983-08-01

    Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

  9. Estudio del balance energético en velocistas de alto rendimiento

    OpenAIRE

    Lorente Gutiérrez, Jesús

    2015-01-01

    El objetivo de este estudio fue evaluar el balance energético en tres atletas de alto rendimiento durante 28 días, que coincidieron con el periodo competitivo de pista cubierta. La ingesta energética fue estudiada a partir de registros alimentarios durante los 28 días. Del mismo modo, el gasto energético fue estimado por tres métodos, mediante registros de actividad durante 28 días, mediante el estudio del ritmo metabólico basal estudiado por calorimetría indirecta, aplicándole el fa...

  10. HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

    International Nuclear Information System (INIS)

    Finch, D.R.

    1965-01-01

    1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

  11. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

  12. Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Eidem, G.R.

    1982-06-01

    About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

  13. A review of the probabilistic safety assessment application to the TR-2 research reactor

    International Nuclear Information System (INIS)

    Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

    2001-01-01

    A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

  14. Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.

    1985-04-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

  15. Efecto de un programa de estiramientos activos en jugadoras de fútbol sala de alto rendimiento

    Directory of Open Access Journals (Sweden)

    Francisco Ayala

    2010-01-01

    Full Text Available El propósito de este estudio fue determinar la progresión de la flexibilidad isquiosural a través del rango de movimiento (ROM de la flexión de cadera antes (línea base de la flexión de cadera, duración 6 semanas, durante (efecto de un programa de estiramientos, duración 8 semanas y después (mantenimiento de la flexibilidad, duración 4 semanas de un programa de estiramientos activos de 8 semanas en jugadoras de alto rendimiento de fútbol sala. Material y método: 10 jugadoras de fútbol sala de alto rendimiento adultas jóvenes completaron este estudio. Un diseño longitudinal, temporal, ininterrumpido de medidas repetidas fue utilizado. Todas las jugadoras llevaron a cabo un programa de estiramientos de 8 semanas. El test unilateral de elevación de la pierna recta fue empleado para evaluar el ROM de la flexión de cadera a las 0, 2, 4 y 6 semanas antes del programa de estiramientos, durante el programa de estiramientos, semanas 2, 4, 6 y 8; y a las 2 y 4 semanas después del cese del programa de estiramientos. Resultados y conclusiones: el análisis de la fase inicial reveló que la línea base del ROM de la flexión de cadera tenía una tendencia irregular, con valores máximos de la flexibilidad isquiosural tanto positivos como negativos. El programa de estiramientos activos de 8 semanas aumentó el ROM de la flexión de cadera un 25,96±8,76%. Cuatro semanas después del cese del programa de estiramientos, el ROM de la flexión de cadera mostró un descenso significativo del 7,9%.

  16. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1986-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity, for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. This existence of significant neutron streaming also explains the high count rate observed with the source range monitors that are located in the TMI-2 reactor cavity. (author)

  17. Degradation of gas-phase trichloroethylene over thin-film TiO2 photocatalyst in multi-modules reactor

    International Nuclear Information System (INIS)

    Kim, Sang Bum; Lee, Jun Yub; Kim, Gyung Soo; Hong, Sung Chang

    2009-01-01

    The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO 2 . A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

  18. Prática pedagógica aos educandos com deficiência intelectual numa escola de ensino fundamental com alto IDEB

    OpenAIRE

    Wilma Carin Silva Porta

    2015-01-01

    No atual contexto da inclusão escolar, em que os educandos com deficiência intelectual constituem a maioria dentre os demais deficientes, indagações sobre a atuação dos professores da sala comum das escolas com alto Índice de Desenvolvimento da Educação Básica (Ideb) com esta população, tornam-se sobremaneira importantes. O presente estudo teve como objeto de análise a prática pedagógica na perspectiva inclusiva, de professores do ciclo I do ensino fundamental, numa escola com alto índice do ...

  19. 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.

  20. 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.

  1. 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

  2. The Interactive Dimension of Communication: The Pragmatics of the Palo Alto Group

    OpenAIRE

    Codruţa Porcar; Cristian Hainic

    2011-01-01

    Our paper proposes to analyze from a semiotic perspective the process of communication as conceived within the Palo Alto Group. We will firstly show that, as a result of the Group's critiques and revisions of the linear or mechanistic theories of communication, new perspectives are brought about for the essential axes of transformation within communication: we do not communicate as from a distinct atom to another, through an isolated channel, but through parts which are equal to the whole, th...

  3. A gas-phase reactor powered by solar energy and ethanol for H2 production

    International Nuclear Information System (INIS)

    Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

    2014-01-01

    In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

  4. 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

  5. Sobre la determinación de la calidad de las escorias de horno alto y de las puzolanas

    Directory of Open Access Journals (Sweden)

    Wittekindt, W.

    1964-06-01

    Full Text Available Not availableEn la molienda de clínker portland con escoria de horno alto para obtener cemento portland de hierro y cemento de horno alto, uno está más seguro de la calidad del clínker que de la escoria. El análisis del clínker y el cálculo de los minerales de clínker, facilitado por este análisis, nos dice ya, en gran parte, si se puede fabricar con este clínker un cemento portland de mayor o menor resistencia, si pueden esperarse buenas resistencias iniciales o si hay que contar más bien con mayores resistencias finales, cual es la resistencia del cemento a los sulfatos, etc.

  6. Busca de estruturas em grandes escalas em altos redshifts

    Science.gov (United States)

    Boris, N. V.; Sodré, L., Jr.; Cypriano, E.

    2003-08-01

    A busca por estruturas em grandes escalas (aglomerados de galáxias, por exemplo) é um ativo tópico de pesquisas hoje em dia, pois a detecção de um único aglomerado em altos redshifts pode por vínculos fortes sobre os modelos cosmológicos. Neste projeto estamos fazendo uma busca de estruturas distantes em campos contendo pares de quasares próximos entre si em z Â3 0.9. Os pares de quasares foram extraídos do catálogo de Véron-Cetty & Véron (2001) e estão sendo observados com os telescópios: 2,2m da University of Hawaii (UH), 2,5m do Observatório de Las Campanas e com o GEMINI. Apresentamos aqui a análise preliminar de um par de quasares observado nos filtros i'(7800 Å) e z'(9500 Å) com o GEMINI. A cor (i'-z') mostrou-se útil para detectar objetos "early-type" em redshifts menores que 1.1. No estudo do par 131046+0006/J131055+0008, com redshift ~ 0.9, o uso deste método possibilitou a detecção de sete objetos candidatos a galáxias "early-type". Num mapa da distribuição projetada dos objetos para 22 escala. Um outro argumento em favor dessa hipótese é que eles obedecem uma relação do tipo Kormendy (raio equivalente X brilho superficial dentro desse raio), como a apresentada pelas galáxias elípticas em z = 0.

  7. Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

    International Nuclear Information System (INIS)

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

    1985-01-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

  8. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  9. Analysis of key hardware factors and countermeasure for restricting 49-2 swimming pool reactor lifetime

    International Nuclear Information System (INIS)

    Zhang Yadong; Guo Yue; Yang Xiao; Wang Yiwei; Wang Zhanwen

    2013-01-01

    Safe operation is the most important factor to determine the lifetime of aged 49-2 swimming pool reactor. In this paper, the hardware factors of lifetime were analyzed, such as the pool concrete aging, corrosion of aluminum container and primary coolant system, and graphite swelling etc., and then the corresponding measures such as surveillance, prevention and maintenance were purposed. The results show that 49-2 swimming pool reactor can continue to operate safely due to that container is safe under 8 degree earthquake, the reactor is safe on flood level of once per millennium, adding dam break, and the ageing condition of primary coolant system and container is acceptable. (authors)

  10. Grave number 121 of the argaric site of Castellón Alto (Galera, Granada

    Directory of Open Access Journals (Sweden)

    Molina, Fernando

    2003-06-01

    Full Text Available A new grave with partly mummified bodies was discovered during fieldwork to prepare the argaric site of Castellón Alto for public visits. Timber slabs and a dry stone wall seal the artificial cave preserving the interior. The human bones belong to one adult and one infant, both with preserved hair and skin fragments. The grave goods comprise several pottery vessels, one dagger, one ax with wooden handle, metal ornaments and fragments of flax and possibly wool.

    Recientes excavaciones en el yacimiento argárico de Castellón Alto con motivo de los trabajos de acondicionamiento para su visita publica han permitido descubrir una sepultura con restos humanos momificados en su interior. La sepultura de tipo covacha se encontraba sellada por tablones de madera y un muro de mampostería. En el interior aparecieron un individuo adulto y un infantil que conservan restos de pelo y piel. El ajuar se compone de varias vasijas cerámicas, un puñal, una azuela con mango de madera y adornos en metal, así como restos de lino y posiblemente lana.

  11. Control Neuroborroso en Red. Aplicación al Proceso de Taladrado de Alto Rendimiento

    Directory of Open Access Journals (Sweden)

    Agustín Gajate

    2009-01-01

    Full Text Available Resumen: Este trabajo muestra el diseño y la implementación de un sistema neuroborroso para el modelado y control en red de un proceso de taladrado de alto rendimiento. El sistema neuroborroso considerado en este estudio es el conocido como Adaptive Network based Fuzzy Inference System (ANFIS, en el que las reglas borrosas se obtienen a partir de datos entrada/salida. Para el diseño del sistema de control se ha elegido el paradigma del control por modelo interno. Los resultados obtenidos son positivos tanto en la simulación como en la aplicación al control en red de la fuerza de corte. Desde el punto de vista técnico, se aumenta la tasa de arranque de material y al mismo tiempo se garantiza un aprovechamiento efectivo de la vida útil de la herramienta de corte. Este buen comportamiento del sistema de control neuroborroso basado en control por modelo interno se ha verificado por medio de varias cifras de mérito. Palabras clave: sistemas neuroborrosos, control por modelo interno, control en red, taladrado de alto rendimiento

  12. The Alto Moxoto Terrain in Eastern Paraiba ('Caldas Brandao Massif')

    International Nuclear Information System (INIS)

    Neves, Benjamim Bley de Brito; Campos Neto, Mario da Costa; Souza, Solange Lucena de; Schmus, William Randall Van; Fernandes, Tania Maria Gomes

    2001-01-01

    The Alto Moxoto Terrane (TAM), at the east of Paraiba State is mostly composed of sheared ortho gneisses, porphyritic granodioritic gneisses and it bears an imbricated sheet of Al-rich (garnet-biotite-sillimanite) gneisses, deeply affected by migmatization phenomena. This litho-structural assemblage is drawing a regional asymmetric anti formal structure, with its axial zone running parallel to the B R-230 highway (E-W trending). It is limited in both, north (Alto Pajeu terrane) and south (Rio Capibaribe terrane) sides by important shear zones, which are feather faults connected with the development of the Pernambuco lineament, to the southwest. The adopted designation of 'terrane' is based upon its singular geological features, in terms of lithological and structural characteristics, Paleoproterozoic in age and sharp limits with the different confining terranes. TAM is here considered as a mega-fragment of the Atlantica Super continent, that was built up by the Paleoproterozoic Collage ('Transamazonian') and that was preserved in the framework of West Gondwana (Brasiliano/Pan African Collage) as a 'terrane'. This terrane shows conspicuous continuity to the far interior of the province, to the southwestern part of Pernambuco State, and so doing, it demonstrates that the former designation of 'Caldas Brandao Massif must be ruled out, as obsolete for many reasons. Geochronological determinations using Rb-Sr, Sm-Nd and U-Pb methods confirm the Paleoproterozoic age of this terrane, with the presence of some Archean protoliths as well as the various degrees of structural reworking and isotopic reseting promoted by the Brasiliano Cycle. This cycle was responsible for some intrusive granites, for most of the general geological features, like usual informal limits and even the present shape of the TAM, a typical reworked 'basement inlier'. (author)

  13. The Alto Paraguay Alkaline Province: petrographic, geochemical and geochronological characteristics

    International Nuclear Information System (INIS)

    Velazquez Fernandez, Victor

    1996-01-01

    The Alto Paraguay Province is located at the border of the State of Mato Grosso do Sul and Paraguay, between the coordinates 21 deg 10 ' to 23 deg 25 ' of Southern latitude and 57 deg 10 ' to 58 deg 00 ' , having the city of Porto Murtinho as the main reference point. The geotectonic domain of the area is governed by the precambric units of the Southern extreme of the Amazonic craton which developed a long and accentuated activity, giving rise to folds and important faults, that in several cases seem to have exerted an effective control of the magmatic manifestations. Radiometric data indicate that the emplacement of the syenitic bodies took place in the Permo-Triassic period, with a major incidence in the interval 260-240 Ma, representing thus, an important phase of alkaline magmatic affinity associated to the Parana Basin which is believed is to be unique, since the other known areas (Central, Amambay and Rio Apa Provinces, Paraguay, Velasco Province, Bolivia) are considerably younger (140-120 Ma). Syenitic rocks from the Alto Paraguay Province show wide variation in the ratio 87 Sr/ 86 Sr (0.703361 - 0.707734). Excluding the Cerro Boggiani rocks (0.703837-0.707734), values for the nepheline syenites (0.703361-0.703672) general lower than those of the other syenites types. Alkaline syenites cover the interval 0.703510- 0.703872, while quartz syenites and syenogranites are 0.704562 and 0.707076, respectively. geologic evidence, in addition to petrographic, geochemical and isotopic (Sr) data, suggest that the syenitic rocks have been derived from an unique mantelic parental liquid, by fractional crystallization and assimilation processes, which are assumed to be occurred during the emplacement of the magma in the crust. (author)

  14. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    additional consideration should be required in nuclear design and fuel treating facilities due to reactivity coefficient being shifted to the plus side, larger neutron yield and increased heat source caused by MA loading. (2) Confirmation of TRU burning reactor core concepts. The core specification of sodium cooled-nitride fueled TRU burning large reactor was designed based on commercial type fast reactor (sodium cooled nitride fueled large fast reactor, 38000 MWt) which was designed in the feasibility studies on commercialized fast reactor cycle system. The composition of MAs from LWR's spent fuel was supposed. MA content in the core fuel is settled to 60 wt% based on the JAERI's design in order to maximize the MA transmutation amount. We need to exchange 25% of core fuel with zirconium hydride (ZrH 1.6 ) to attain Doppler coefficient being equivalent to that of the conventional type commercial fast reactor loaded 5 wt% MA. Furthermore, this reactor could transmute MAs produced in forty-eight sodium cooled nitride fueled large fast reactors generating the same output. In order to investigate the dependency of MA transmutation characteristics on the reactor output, 1200 MWt TRU burning middle or small reactor core concept was designed. This core was settled by reducing the number of core fuel assemblies from that of TRU burning large reactor designed above. MA transmutation rate in this core is smaller than that in the TRU burning large reactor core because the neutron flux of this core becomes smaller than that of the TRU burning large reactor core due to the higher Pu enrichment. (3) Comparison between TRU burning reactor and conventional type commercial fast reactor. MA transmutation and nuclear characteristics of the sodium cooled nitride fuel commercial type fast reactor loaded 5 wt%MA were evaluated and compared with those of TRU burning large reactor designed in (2). The commercial type fast reactor could only transmute MAs produced in seven sodium cooled nitride

  15. Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

    1997-07-01

    Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

  16. Altos custos financeiros do trauma vascular

    Directory of Open Access Journals (Sweden)

    Ricardo Costa-Val

    Full Text Available OBJETIVO: Demonstrar o custo e impacto financeiro referente à primeira abordagem cirúrgica das lesões vasculares em pacientes admitidos no Hospital João XXIII/FHEMIG, entre os anos de 2004 a 2006. MéTODOS: Trata-se de um estudo com aprovação ética, retrospectivo, de coorte e descritivo realizado a partir da auditoria de contas hospitalares referentes a 70 prontuários catalogados pelo Serviço de Trauma Cardiovascular. RESULTADOS: Cinco (7,14% prontuários foram excluídos por má qualidade técnica. O valor monetário repassado pelo Sistema Único de Saúde e pelo setor privado foram de R$ 103.614,96 (US$ 60.949,97 e de R$ 185.888,21 (US$ 109.346,0, respectivamente, implicando em defasagem potencial de 44%. Houve correlação direta entre custos e topografia anatômica das lesões e exponencial em relação às variáveis hemoderivados e próteses vasculares. CONCLUSÃO: Este estudo corrobora os altos custos do trauma vascular e fortalece a importância da auditoria de contas para as tomadas de decisões médicas.

  17. Análisis de la matutinidad-vespertinidad en jóvenes atletas de alto rendimiento

    Directory of Open Access Journals (Sweden)

    Alejo Sebastián García-Naveira Vaamonde

    2015-01-01

    Full Text Available El objetivo de este trabajo fue estudiar la relación entre la matutinidad-vespertinidad, la edad, el sexo, la ansiedad rasgo y la modalidad deportiva en depor- tistas adolescentes. La muestra estaba formada por 102 jóvenes atletas españoles de alto rendimiento (54 mujeres y 48 hombres con una edad entre los 14 y 17 años. Se midió la matutinidad-vespertinidad mediante la Escala Compuesta de Matutinidad-Vespertinidad (CS y la ansiedad rasgo mediante el Inventario de Ansie- dad Estado-Rasgo (STAI. Los resultados indican que no existe relación entre el cronotipo, la edad, el sexo y la ansiedad de los deportistas, mientras que estos son más matutinos que la población general de adolescentes y los velocistas/vallistas son más vespertinos que el resto de modalidades. Se concluye que la práctica deportiva de alto rendimiento puede que sea un Zeitzbergs ex- terno que modifica aspectos psicológicos, fisiológicos y bioquímicos de los jóvenes.

  18. On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1994-01-01

    In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

  19. Organización socioeconómica y territorial en la región del Alto Lerma, Estado de México

    Directory of Open Access Journals (Sweden)

    Estela Orozco Hernández

    2012-02-01

    Full Text Available Este trabajo muestra que la región del Alto Lerma es un espacio de organización compleja, donde se entrelazan procesos sociales y territoriales diversos, representados por la existencia de estructuras agrarias, urbanas e industriales. Cada una de estas estructuras tiene necesidades e intereses que definen las formas de apropiación, control y producción del espacio regional y, por lo tanto, constituyen factores determinantes de la configuración socioterritorial del Alto Lerma. Se analiza la información estadística oficial, así como los trabajos disponibles, desde una perspectiva hipotético-deductiva.

  20. Caracterización de las gestantes de alto riesgo obstétrico (ARO en el departamento de Sucre (Colombia, 2015

    Directory of Open Access Journals (Sweden)

    Judith Martínez Royert

    2016-01-01

    Full Text Available Objetivo: Caracterizar las gestantes de Alto Riesgo Obstétrico ( ARO que acuden a una IPS pública en el departamento de Sucre, Colombia (periodo enero, febrero y marzo de 2015. Material y métodos: Estudio cuantitativo, descriptivo. La muestra la conformaron 123 gestantes ARO . Se utilizó un instrumento elaborado por las investigadoras; se sometió a validez de constructo y contenido, y análisis de consistencia interna mediante el alfa de Cronbach. Resultados: El 13,18 % de las gestantes eran menores de 18 años; 38,2% procedentes de la capital y 19,5 % de la región del San Jorge; 66 % no manifestaron antecedentes patológicos; 13,8 % presentaron complicaciones de amenaza de aborto o de parto pretérmino; 37 % eran nulípara; 20.3 % tenían cesárea anterior; 22.8 % sufrieron abortos; 54.5 % manifestaron tensión emocional y mal humor; 82.9% no programaron el embarazo; 24 % con periodo intergenésico de 1 año; 55.3 % (68 gestantes se encontraban entre la semana 30 y 40 de gestación al momento de participar en el estudio. Conclusiones: La subregión de la Sabana y San Jorge fueron las que presentaron mayor número de gestantes de alto riesgo. Entre las patologías preexistentes más frecuentes se encontró anemias y migrañas, así como las del sistema endocrino y respiratorio. Esta investigación servirá como referente para proporcionar conocimiento respecto al perfil de las gestantes de alto riesgo en Sucre, para que los profesionales involucrados en su atención desempeñen un rol que permita contribuir al control y prevención de las complicaciones en ellas y en la reducción significativa de la mortalidad materna.

  1. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik

    2015-01-01

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized

  2. VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

    International Nuclear Information System (INIS)

    Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

    1980-01-01

    1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

  3. Entre aromas de incienso y pólvora : Los Altos de Jalisco, México, 1917-1940

    NARCIS (Netherlands)

    López Ulloa, José Luis

    2008-01-01

    This thesis focuses on the Mexican Revolution and on the opposition strategies followed by the opponents of the revolutionary regime who lived in the region known as Los Altos de Jalisco. In this particular region, the Catholic population, supported by the Clergy, was in constant conflict with the

  4. Tolerancia a la violencia de pareja en tres historias de vida de mujeres de estrato económico alto de Lima

    OpenAIRE

    Mujica, Jaris; Bedoya, Silvana

    2017-01-01

    Objetivo: esta investigación analiza la presencia de violencia física y psicológica ejercidas por la pareja (varón) a una víctima (mujer) en el estrato económico alto de Lima. Método: se recolectaron los datos a través de la técnica “historia de vida” en un registro profundo de los discursos de tres mujeres adultas del estrato económico alto. Resultados: los resultados ratifican resultados de la literatura precedente (concentrada en sectores de medios y bajos recursos económicos) en las forma...

  5. Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

    2013-07-01

    In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

  6. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  7. Nuclear structure of neutron rich gallium, germanium and arsenic around N=50 and development of a laser ion source at ALTO

    International Nuclear Information System (INIS)

    Tastet, B.

    2011-01-01

    During this thesis, we have studied β decays of gallium's nuclei around N=50 and prepared a laser ionization source at ALTO.The production of exotic isotopes has brought new beam production challenges. The one addressed here relates to the elimination of isobar contaminants that create background for experiments. To address this issue a laser ionization source has been developed at ALTO. Copper has been chosen to be the first element to be ionized for physical interests and to compare the results of the laser ionization source with the ones at others facilities. A laser setup has been installed and optimized in order to ionize selectively the atoms of copper produced for experiments. After the optimization, a test of ionization of stable-copper was performed. This test has shown us that the laser system is able to successfully ionize atoms of copper.The studies of the region of the neutron-rich nuclei around N=50 are still to complete. 79,80,82,83,84,85 Ga has been produced using photo-nuclear reactions at the experimental area of the on-line PARRNe mass-separator operating with the ALTO facility. The fission fragments are produced at the interaction of the 50 MeV electron beam delivered by the ALTO linear accelerator with a thick target of uranium in a standard UC x form. The oven is connected to a W ionizer heated up to 2000 C degrees that selectively ionizes alkalis but also elements with low ionization potentials such as Ga. The ions are accelerated through 30 kV and magnetically mass-separated before being implanted on a mylar tape close to the detection setup, so that this system allows us to study β and β-n decays of 79,80,82,83,84,85 Ga.The data analysis have produced new results concerning the decays of 80 Ga, 84 Ga and 84 Ge. For 80 Ga, the existence of an isomeric state has been confirmed and two different half-lives were measured for the ground state and the isomer. Furthermore, the analysis of 84 Ga decay confirmed two states and allowed us to

  8. Proceedings of the international topical meeting on advanced reactors safety: Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

  9. Conservation and valorisation of Giovanni Boccaccio’s house museum in Certaldo Alto

    Directory of Open Access Journals (Sweden)

    Massimo Gennari

    2012-04-01

    Full Text Available The architectural restoration and functional redevelopment of Boccacio’s house in Certaldo Alto (Florence, has been carried out between 2006 and 2007 and finalized in 2011, including the reconstruction of the garden next to the house. The program which has been characterized by a strong civic, social and cultural involvement, is lead by the Ente Nazionale G. Boccaccio in partnership whit the local administration and aims at contributing to the valorization of the historical, architectural and cultural heritage of the historic center of Certaldo Alto. Valorization here is intended as a functional integration and synergy between predominantly cultural activities. The aim is to achieve the best results in terms of social development as well as intellectual growth within a virtuous economy, and therefore the construction of a complementary model for cultural assets in general. A model where the single cultural elements (museums, libraries, workrooms, exhibitions, auditorium, etc.. represent only the intersections of a wider net system established through the process of communication and exchange with the institutions, publics or privates, that operate in the sectors of research, experimentation, education and information. This means that the management of cultural assets will now aim mainly at the interaction between its components and nationals as well as international structures of education and research, institutes for the social and economical development and innovative business structures in the fields of communication and cultural and sustainable tourism. This establishes an additional value of his still underestimated significance.

  10. 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)

  11. Application of 2DOF controller for reactor power control. Verification by numerical simulation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1996-09-01

    In this report the usefulness of the two degree of freedom (2DOF) control is discussed to improve the reference response characteristics and robustness for reactor power control system. The 2DOF controller consists of feedforward and feedback elements. The feedforward element was designed by model matching method and the feedback element by solving the mixed sensitivity problem of H ∞ control. The 2DOF control gives good performance in both reference response and robustness to disturbance and plant perturbation. The simulation of reactor power control was performed by digitizing the 2DOF controller with the digital control periods of 10[msec]. It is found that the control period of 10[msec] is enough not to make degradation of the control performance by digitizing. (author)

  12. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2007-01-01

    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

  13. 77 FR 58203 - AER Energy Resources, Inc.; Alto Group Holdings, Inc.; Bizrocket.Com Inc.; Fox Petroleum, Inc...

    Science.gov (United States)

    2012-09-19

    ... SECURITIES AND EXCHANGE COMMISSION [File No. 500-1] AER Energy Resources, Inc.; Alto Group Holdings, Inc.; Bizrocket.Com Inc.; Fox Petroleum, Inc.; Geopulse Explorations Inc.; Global Technologies... accuracy of press releases concerning the company's revenues. 4. Fox Petroleum, Inc. is a Nevada...

  14. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

  15. Degradation of gas-phase trichloroethylene over thin-film TiO{sub 2} photocatalyst in multi-modules reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Bum [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Lee, Jun Yub, E-mail: ljy02191@hanafos.com [Power Engineering Research Institute, Korea Power Engineering Company, Inc. (Korea, Republic of); Kim, Gyung Soo [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Hong, Sung Chang [Department of Environmental Engineering, Kyonggi University (Korea, Republic of)

    2009-07-30

    The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO{sub 2}. A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

  16. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  17. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  18. 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

  19. Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

    CERN Document Server

    Chiba, G

    2002-01-01

    It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

  20. 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

  1. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    International Nuclear Information System (INIS)

    Potapov, I.A.; Serebrov, A.P.

    2001-01-01

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  2. Un centro ceremonial formativo en el Alto Piura

    Directory of Open Access Journals (Sweden)

    1989-01-01

    estilos diversos, cuya evolución temporal se puede seguir, tanto del punto de vista de las formas y técnicas decorativas como de la iconografía, abundante y diversificada. Estos datos, nuevos para la región, así como el análisis comparativo de los vestigios materiales, comprueban la existencia de contactos y relaciones con las demás zonas cercanas y particularmente la integración del Alto Piura a los sistemas ideológicos y religiosos más sureños. La implantación del sitio podría estar ligada a su ubicación geográfica, en el cruce de una vía de intercambio entre poblaciones costeras, andinas y selváticas, y de un camino norte-sur facilitando los contactos entre la costa norte peruana y la costa y los Andes ecuatorianos. Su ocupación parece testimoniar una situación original -caracterizada por la presencia de representantes de varias tradiciones culturales- que se mantendrá e igualmente singularizará este sector del Alto Piura durante las épocas posteriores. Research carried out since 1986 on the archaeological site of Cerro Ñañañique (Chulucanas, department of Piura has allowed the identification and description of a ceremonial complex, built and occupied between the Xth and the Vth centuries B.C.. Several phases of edification and widening of the complex have been recognized according to a general U shape plan. Several ceramic traditions of various styles and origins can be found contemporarily on the site, and their temporal evolution can be identified and followed from these points of view: forms, decorative techniques and iconography which is abundant and diverse. The data, new for this study area, and the comparative analysis of material remains demonstrate the existence of relations with nearby zones and, particularly, the integration of Alto Piura in Southern ideological and religious systems. The establishment of the site might be related to its geographic location, at the crossing of an exchange route between coastal, andean and amazonian

  3. Manual de auditoría interna para Instituto de Altos Estudios Nacionales

    OpenAIRE

    Bungacho Lamar, Fredy, Dr.

    2010-01-01

    La Ley Orgánica de Administración Financiera y Control responsabiliza a cada institución del Estado la implementación y aplicación del Sistema de Control Interno con la finalidad de precautelar los recursos públicos. Este trabajo investigativo sirve como guía para unificar los procedimientos, en la ejecución de las Auditoría de los profesionales que integren la Unidad de auditoría Interna del Instituto de altos Estudios Nacionales. Este Manual de Auditoría Interna, se compone de 6 capí...

  4. Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

    International Nuclear Information System (INIS)

    1978-07-01

    The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

  5. 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

  6. Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Jin, Yoon-Su; Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2014-09-15

    Highlights: • The authors designed and fabricated a D-shape coil based toroid-type HTS DC reactor using 2G GdBCO HTS wires. • The toroid-type magnet consisted of 30 D-shape double pancake coil (DDC)s. The total length of the wire was 2.32 km. • The conduction cooling method was adopted for reactor magnet cooling. • The maximum cooling temperature of reactor magnet is 5.5 K. • The inductance was 408 mH in the steady-state condition (300 A operating). - Abstract: This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

  7. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  8. Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

    International Nuclear Information System (INIS)

    Kim, Tae H.; Yang, Chae Y.

    2013-01-01

    The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

  9. Mass transfer of ammonia escape and CO2 absorption in CO2 capture using ammonia solution in bubbling reactor

    International Nuclear Information System (INIS)

    Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing

    2016-01-01

    Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.

  10. SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

    Science.gov (United States)

    Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

    2016-04-01

    Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

  11. A novel condensation reactor for efficient CO2 to methanol conversion for storage of renewable electric energy

    NARCIS (Netherlands)

    Bos, Martin Johan; Brilman, Derk Willem Frederik

    2015-01-01

    A novel reactor design for the conversion of CO2 and H2 to methanol is developed. The conversion limitations because of thermodynamic equilibrium are bypassed via in situ condensation of a water/methanol mixture. Two temperatures zones inside the reactor ensure optimal catalyst activity (high

  12. Quality assurance in the project of RECH-2 research reactor

    International Nuclear Information System (INIS)

    Goycolea Donoso, C.; Nino de Zepeda Schele, A.

    1989-01-01

    The implantation of a Quality Assurance Program for the design, supply, construction, installation, and testing of the RECH-2 research reactor, is described in this paper. The obtained results, demonstrate that a Quality Assurance Program constitutes a suitable mean to assure that the installation complies with the safety and reliability requirements. (author)

  13. Ethnobotanical and phytomedicinal knowledge in a long-history protected area, the Abruzzo, Lazio and Molise National Park (Italian Apennines).

    Science.gov (United States)

    Idolo, Marisa; Motti, Riccardo; Mazzoleni, Stefano

    2010-02-03

    This study reports on the ethnobotanical and phytomedical knowledge in one of the oldest European Parks, the Abruzzo, Lazio and Molise National Park (Central Italy). We selected this area because we judged the long history of nature preservation as an added value potentially encouraging the survival of uses possibly lost elsewhere. In all, we interviewed 60 key informants (30 men and 30 women) selected among those who, for their current or past occupation or specific interests, were most likely to report accurately on traditional use of plants. The average age of informants was 65 years (range 27-102 years). The ethnobotanical inventory we obtained included 145 taxa from 57 families, corresponding to 435 use-reports: 257 referred to medical applications, 112 to food, 29 to craft plants for domestic uses, 25 to veterinary applications, 6 to harvesting for trade and another 6 to animal food. The most common therapeutic uses in the folk tradition are those that are more easily prepared and/or administered such as external applications of fresh or dried plants, and decoctions. Of 90 species used for medical applications, key informants reported on 181 different uses, 136 of which known to have actual pharmacological properties. Of the uses recorded, 76 (42%) concern external applications, especially to treat wounds. Medical applications accounted for most current uses. Only 24% of the uses we recorded still occur in people's everyday life. Species no longer used include dye plants (Fraxinus ornus, Rubia tinctorum, Scabiosa purpurea, Rhus coriaria and Isatis tinctoria) and plants once employed during pregnancy, for parturition, nursing, abortion (Asplenium trichomanes, Ecballium elaterium, Juniperus sabina and Taxus baccata) or old magical practices (Rosa canina). Our study remarked the relationship existing between the high plant diversity recorded in this biodiversity hotspot of central Apennines and the rich ethnobotanical knowledge. The presence of some very

  14. Set of rules SOR 2 licensing of nuclear reactors

    International Nuclear Information System (INIS)

    1976-05-01

    This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

  15. Prevalencia de genotipos del virus del papiloma humano de alto riesgo no vacunables dentro del programa de Detección Precoz de Cáncer de Cérvix en Cantabria

    Directory of Open Access Journals (Sweden)

    María Paz-Zulueta

    2016-06-01

    Conclusiones: Atendiendo al alto porcentaje de VPH de alto riesgo oncogénico no vacunable, habría que replantear la estrategia de prevención en la población, que podría tener una falsa sensación de protección.

  16. Measurement of thermal conductivity of sintered UO{sub 2} in the reactor; Merenje toplotne provodljivosti sinterovanog UO{sub 2} u reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Katanic, J; Stevanovic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1965-10-15

    Thermal conductivity is considered one of the fundamental properties of sintered UO{sub 2} fuel. Samples should be tested under real core conditions. This paper covers the methods and instruments for thermal conductivity measurement of UO{sub 2} samples in the reactor core, measurements outside the core under conditions similar to those in the core and outside the core after irradiation. Fuel samples are placed in capsules for irradiation in the reactor in-core loops.

  17. Geomechanical Analysis of Underground Coal Gasification Reactor Cool Down for Subsequent CO2 Storage

    Science.gov (United States)

    Sarhosis, Vasilis; Yang, Dongmin; Kempka, Thomas; Sheng, Yong

    2013-04-01

    Underground coal gasification (UCG) is an efficient method for the conversion of conventionally unmineable coal resources into energy and feedstock. If the UCG process is combined with the subsequent storage of process CO2 in the former UCG reactors, a near-zero carbon emission energy source can be realised. This study aims to present the development of a computational model to simulate the cooling process of UCG reactors in abandonment to decrease the initial high temperature of more than 400 °C to a level where extensive CO2 volume expansion due to temperature changes can be significantly reduced during the time of CO2 injection. Furthermore, we predict the cool down temperature conditions with and without water flushing. A state of the art coupled thermal-mechanical model was developed using the finite element software ABAQUS to predict the cavity growth and the resulting surface subsidence. In addition, the multi-physics computational software COMSOL was employed to simulate the cavity cool down process which is of uttermost relevance for CO2 storage in the former UCG reactors. For that purpose, we simulated fluid flow, thermal conduction as well as thermal convection processes between fluid (water and CO2) and solid represented by coal and surrounding rocks. Material properties for rocks and coal were obtained from extant literature sources and geomechanical testings which were carried out on samples derived from a prospective demonstration site in Bulgaria. The analysis of results showed that the numerical models developed allowed for the determination of the UCG reactor growth, roof spalling, surface subsidence and heat propagation during the UCG process and the subsequent CO2 storage. It is anticipated that the results of this study can support optimisation of the preparation procedure for CO2 storage in former UCG reactors. The proposed scheme was discussed so far, but not validated by a coupled numerical analysis and if proved to be applicable it could

  18. Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Wu Yuanyuan; Liu Tiancai; Sun Wei

    2012-01-01

    The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

  19. Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

    International Nuclear Information System (INIS)

    Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

    1979-01-01

    Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

  20. 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.)

  1. Revision of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

    To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

  2. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  3. N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

    Czech Academy of Sciences Publication Activity Database

    Obalová, L.; Jirátová, Květa; Karásková, K.; Chromčáková, Ž.

    2012-01-01

    Roč. 191, č. 1 (2012), s. 116-120 ISSN 0920-5861 R&D Projects: GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : N2O * catalytic decomposition * fixed bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.980, year: 2012

  4. 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

  5. Alteration of installation of reactors (alteration of No.1 and No.2 reactor facilities) in Oi Power Station, Kansai Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1984-01-01

    The Nuclear Safety Commission reported to the Minister of International Trade and Industry on October 27, 1983, that the technical capability was recognized to be adequate, and the safety after the alteration of the installation of reactors was judged to be ensured. At the time of deliberation, the guidelines for examining the safety design and safety evaluation of LWR facilities for power generation were used. Regarding the change of the degree of enrichment of replacement fuel from 3.2 to 3.4 wt.%, the limiting conditions are satisfied in the replacement core, and the nuclear design is appropriate. Eight test fuel assemblies using UO 2 pellets containing gadolinia are charged in the core of No.2 reactor, and the irradiation of two cycles is carried out. As the result of the safety examination regarding this test, the propriety of the nuclear design and mechanical design of the test fuel assemblies was confirmed. This alteration does not exert influence on the result of safety analysis made so far. This report was decided by the Committee on Examination of Reactor Safety based on the conclusion of No.26 subcommittee. (Kako, I.)

  6. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    International Nuclear Information System (INIS)

    Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

    2008-01-01

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  7. Plasma-catalyst hybrid reactor with CeO2/γ-Al2O3 for benzene decomposition with synergetic effect and nano particle by-product reduction.

    Science.gov (United States)

    Mao, Lingai; Chen, Zhizong; Wu, Xinyue; Tang, Xiujuan; Yao, Shuiliang; Zhang, Xuming; Jiang, Boqiong; Han, Jingyi; Wu, Zuliang; Lu, Hao; Nozaki, Tomohiro

    2018-04-05

    A dielectric barrier discharge (DBD) catalyst hybrid reactor with CeO 2 /γ-Al 2 O 3 catalyst balls was investigated for benzene decomposition at atmospheric pressure and 30 °C. At an energy density of 37-40 J/L, benzene decomposition was as high as 92.5% when using the hybrid reactor with 5.0wt%CeO 2 /γ-Al 2 O 3 ; while it was 10%-20% when using a normal DBD reactor without a catalyst. Benzene decomposition using the hybrid reactor was almost the same as that using an O 3 catalyst reactor with the same CeO 2 /γ-Al 2 O 3 catalyst, indicating that O 3 plays a key role in the benzene decomposition. Fourier transform infrared spectroscopy analysis showed that O 3 adsorption on CeO 2 /γ-Al 2 O 3 promotes the production of adsorbed O 2 - and O 2 2‒ , which contribute benzene decomposition over heterogeneous catalysts. Nano particles as by-products (phenol and 1,4-benzoquinone) from benzene decomposition can be significantly reduced using the CeO 2 /γ-Al 2 O 3 catalyst. H 2 O inhibits benzene decomposition; however, it improves CO 2 selectivity. The deactivated CeO 2 /γ-Al 2 O 3 catalyst can be regenerated by performing discharges at 100 °C and 192-204 J/L. The decomposition mechanism of benzene over CeO 2 /γ-Al 2 O 3 catalyst was proposed. Copyright © 2017 Elsevier B.V. All rights reserved.

  8. COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1994-03-01

    The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

  9. Ethanol production by immobilized yeast and its CO2 gas effects on a packed bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, G M; Choi, C Y; Choi, Y D; Han, M H

    1982-10-01

    Immobilised yeast trapped in an alginate matrix demonstrated maximum activity at 30 degrees C and showed no pH effect between 3 and 7. Substrate inhibition was observed at glucose concentrations above 8% but the immobilised cells retained 70% of their maximum activity at 20% glucose concentration. The operation stability of immobilised cells was lower in simple glucose solution than in the activation medium in which only 20% of the activity was lost after 10 days operation. Inactivated immobilised yeast beads were reactivated by incubation in activation medium without a significant increase in cell numbers in a bead. During the operation of the immobilised yeast in a packed bed reactor, CO/sub 2/ gas accumulation adversely affected the reactor performance. An ideal plus flow reactor, not taking into account the formation of CO/sub 2/ gas bubbles and the presence of mass trasnfer resistance, was simulated using a kinetic model for the production of ethanol and the simulation results were compared with the actual reactor performance to determine the CO/sub 2/ gas effect, quantitatively. Up to 45% of the substrate conversion was lost due to the accumulation of CO/sub 2/ gas bubbles in all cases. (Refs. 21).

  10. Energy saving in the pig iron production in the blast furnace no. 5; Ahorro de energia en la produccion de arrabio en el alto horno No. 5

    Energy Technology Data Exchange (ETDEWEB)

    Gil Diaz, Ricardo A; J Quiroz, Francisco; Rodriguez, Rita Patricia; Banuelos Garza, Yolanda [Altos Hornos de Mexico, S. A., Coahuila (Mexico)

    1994-12-31

    Altos Hornos de Mexico (AHMSA) is an iron and steel industry integrated to Grupo Acereros del Norte in Monclova, in the Coahuila state. With an a installed capacity of 3.1 millions of tons per annum o liquid steel. In its installations, AHMSA has the highest capacity blast furnace installed in Mexico, blast furnace No. 5, that has a useful volume of 2,163 cubic meters, designed to produce 4,800 tons of pig iron per day. The basic goal to achieve in the operations involved in the production of steel through the pig iron production in the blast furnace, is the hot metal production at the lowest attainable cost within the quality requirements specified by the steel makers. The most important criterion for the recognition of the attained success is the fuel consumption per ton of pig iron produced, with coke as the main fuel fed to the blast furnace and therefore of the greatest impact on the final product cost. AHMSA contemplated within its strategic plan, the reduction in the production of its coking plants derived from the natural aging of its furnaces, consequently it is pending the shortage of coke for productions higher than 2.6 MMT of liquid iron. In response to this, and faced to the true need of diminishing the production costs in the process of making pig iron, new practices have been implemented in the use of complementary fuels to partially substitute the metallurgical coke as an energy source for the blast furnace process. The use of natural gas, fuel oil and the gradual increase of the temperature of hot blow, have strongly impacted the metallurgical coke consumption, lowering it considerably and diminishing the costs per ton of pig iron in blast furnace No. 5. Another important issue, is the utilization of coke fines resulting form the sieving of the same, directly fed to the furnace load. This practice reduced the coke consumption, and most of all, the output of our coking plants was increased on being utilized at the maximum coke production

  11. Energy saving in the pig iron production in the blast furnace no. 5; Ahorro de energia en la produccion de arrabio en el alto horno No. 5

    Energy Technology Data Exchange (ETDEWEB)

    Gil Diaz, Ricardo A.; J Quiroz, Francisco; Rodriguez, Rita Patricia; Banuelos Garza, Yolanda [Altos Hornos de Mexico, S. A., Coahuila (Mexico)

    1993-12-31

    Altos Hornos de Mexico (AHMSA) is an iron and steel industry integrated to Grupo Acereros del Norte in Monclova, in the Coahuila state. With an a installed capacity of 3.1 millions of tons per annum o liquid steel. In its installations, AHMSA has the highest capacity blast furnace installed in Mexico, blast furnace No. 5, that has a useful volume of 2,163 cubic meters, designed to produce 4,800 tons of pig iron per day. The basic goal to achieve in the operations involved in the production of steel through the pig iron production in the blast furnace, is the hot metal production at the lowest attainable cost within the quality requirements specified by the steel makers. The most important criterion for the recognition of the attained success is the fuel consumption per ton of pig iron produced, with coke as the main fuel fed to the blast furnace and therefore of the greatest impact on the final product cost. AHMSA contemplated within its strategic plan, the reduction in the production of its coking plants derived from the natural aging of its furnaces, consequently it is pending the shortage of coke for productions higher than 2.6 MMT of liquid iron. In response to this, and faced to the true need of diminishing the production costs in the process of making pig iron, new practices have been implemented in the use of complementary fuels to partially substitute the metallurgical coke as an energy source for the blast furnace process. The use of natural gas, fuel oil and the gradual increase of the temperature of hot blow, have strongly impacted the metallurgical coke consumption, lowering it considerably and diminishing the costs per ton of pig iron in blast furnace No. 5. Another important issue, is the utilization of coke fines resulting form the sieving of the same, directly fed to the furnace load. This practice reduced the coke consumption, and most of all, the output of our coking plants was increased on being utilized at the maximum coke production

  12. Caracterización de carbones para la inyección por toberas en el horno alto

    Directory of Open Access Journals (Sweden)

    Babich, A.

    1998-05-01

    Full Text Available The efficiency of blast furnace operation with pulverized coal injection (PCI by tuyeres is determined by the composition and properties of the used coals and by the quality of the ferrous burden and coke. A study in thermobalance of coals to be injected by tuyeres is carried out, and the softening and melting temperatures of coals ash are determined. The coal performance and its influence in the blast furnace operation is estimated.

    La eficacia de la operación del horno alto con inyección de carbón pulverizado (ICP por toberas, está determinada por la composición y propiedades de los carbones utilizados y por la calidad de la carga férrea y del coque. Se realiza el estudio en termobalanza de carbones destinados a la inyección por toberas y se determinan las temperaturas de reblandecimiento y fusión de la ceniza de estos carbones. Se estima el comportamiento de los carbones y su influencia en la operación del horno alto.

  13. O suicídio Tikúna no Alto Solimões: uma expressão de conflitos

    Directory of Open Access Journals (Sweden)

    Regina M. de Carvalho Erthal

    Full Text Available O objetivo deste trabalho é buscar um entendimento a respeito da ocorrência de suicídios entre os índios Tikúna do Alto Solimões (Amazonas, um objeto de difícil aproximação e que aponta para a necessidade de abordagem interdisciplinar. A etnografia realizada preocupou-se em captar a vinculação entre os eventos de suicídio da última década com a exacerbação dos confrontos entre diferentes grupos faccionais que atualizam, em outro contexto histórico, os mecanismos de resolução de conflitos próprios das antigas malocas. Na base desses confrontos está o abandono a que tal população tem sido submetida pelos órgãos responsáveis pela definição e implementação das políticas públicas para as populações indígenas, com especial destaque para a falência do modelo de assistência proposto para a área do Alto Solimões.

  14. Autopia do edifício alto" verde" e a criação de uma nova geração de ícones do desempenho ambiental

    Directory of Open Access Journals (Sweden)

    Erica Mitie Umakoshi

    2009-12-01

    Full Text Available A desordem urbana que tomou conta das cidades da Revolução Industrial, como Londres e Paris, no final do século 19, levou o arquiteto urbanista Le Corbusier a propor um total redesenho urbano, no qual o edifício alto seria o principal elemento de projeto. Nos anos 60, verifica-se o surgimento de projetos visionários baseados nos grandes avanços tecnológicos e voltados às questões habitacionais das grandes cidades, incluindo megaestruturas e edifícios altos. Nesse período, três grupos se destacaram no que tange às inovações do projeto de arquitetura e urbanismo: o Archigram inglês, os Metabolistas japoneses e o Grupo Francês, que criaram uma série de utopias para o tema do edifício alto. Paralelamente, a tipologia do edifício alto de escritórios crescia com o desenvolvimento econômico de importantes cidades do cenário internacional, como Nova York e Londres. À frente, diante da crise energética e ambiental dos anos 70, destacou-se uma nova preocupação na arquitetura mundial: os edifícios deveriam consumir menos energia e serem ambientalmente mais responsáveis. Com isso, surgem utopias que passam a questionar os modelos convencionais, incluindo propostas para o edifício alto" verde". Nos anos 90, o arquiteto malaio Ken Yeang se torna uma referência internacional no tema do" Edifício Alto Ecológico", cujas idéias se baseiam em uma nova estética de projeto: o intenso uso da vegetação, da iluminação e ventilação natural, dentre outras estratégias bioclimáticas, em busca do conforto ambiental. Mais recentemente, sua arquitetura foi reconhecida pelas idéias de" paisagismo vertical" e" urbanismo verde". Ao lado da utopia do edifício verde, exemplos construídos em cidades européias, desde os anos 90, clamam estar definindo as bases arquitetônicas e tecnológicas de uma nova real geração de edifícios mais ecológicos. Os projetos utópicos desenvolvidos ao longo da história vêm exercendo, no que toca

  15. Power Quality Problems Mitigation using Dynamic Voltage Restorer in Egypt Thermal Research Reactor (ETRR-2)

    International Nuclear Information System (INIS)

    Kandil, T.; Ayad, N.M.; Abdel Haleam, A.; Mahmoud, M.

    2013-01-01

    Egypt thermal research reactor (ETRR-2) was subjected to several Power Quality Problems such as voltage sags/swells, harmonics distortion, and short interruption. ETRR-2 encompasses a wide range of loads which are very sensitive to voltage variations and this leads to several unplanned shutdowns of the reactor due to trigger of the Reactor Protection System (RPS). The Dynamic Voltage Restorer (DVR) has recently been introduced to protect sensitive loads from voltage sags and other voltage disturbances. It is considered as one of the most efficient and effective solution. Its appeal includes smaller size and fast dynamic response to the disturbance. This paper describes a proposal of a DVR to improve power quality in ETRR-2 electrical distribution systems . The control of the compensation voltage is based on d-q-o algorithm. Simulation is carried out by Matlab/Simulink to verify the performance of the proposed method

  16. Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

    Energy Technology Data Exchange (ETDEWEB)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-07-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  17. Sistema de alcantarillado por vacío en la Avenida de Valencia y adyacentes de Santa Pola

    OpenAIRE

    Navarro Brotóns, Manuel

    2014-01-01

    Sustitución de sistema de alcantarillado común por el tipo por vacío para eliminar problemas endémicos de pueblos con alto nivel freático marino que transmiten a la EDAR, afectando el alto contenido en sales al reactor biológico.

  18. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  19. Study of the obtainment of Mo_2C by gas-solid reaction in a fixed and rotary bed reactor

    International Nuclear Information System (INIS)

    Araujo, C.P.B. de; Souza, C.P. de; Souto, M.V.M.; Barbosa, C.M.; Frota, A.V.V.M.

    2016-01-01

    Carbides' synthesis via gas-solid reaction overcomes many of the difficulties found in other processes, requiring lower temperatures and reaction times than traditional metallurgic routes, for example. In carbides' synthesis in fixed bed reactors (FB) the solid precursor is permeated by the reducing/carburizing gas stream forming a packed bed without mobility. The use of a rotary kiln reactor (RK) adds a mixing character to this process, changing its fluid-particle dynamics. In this work ammonium molybdate was subjected to carbo-reduction reaction (CH4 / H2) in both reactors under the same gas flow (15L / h) and temperature (660 ° C) for 180 minutes. Complete conversion was observed Mo2C (dp = 18.9nm modal particles sizes' distribution) in the fixed bed reactor. In the RK reactor this conversion was only partial (∼ 40%) and Mo2C and MoO3 (34nm dp = bimodal) could be observed on the produced XRD pattern. Partial conversion was attributed to the need to use higher solids loading in the reactor CR (50% higher) to avoid solids to centrifuge. (author)

  20. 2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

    International Nuclear Information System (INIS)

    Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

    1969-01-01

    1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

  1. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  2. Operating reactors licensing actions summary. Vol. 4, No. 2

    International Nuclear Information System (INIS)

    1984-04-01

    This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  3. Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.

    2012-01-01

    Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones

  4. Assessing impact of hunting mammals in Alto Itaya river basin, Peruvian Amazon

    OpenAIRE

    Aquino, Rolando; Terrones, C.; Navarro, R.; Terrones, Wagner

    2013-01-01

    En el presente trabajo se informa sobre la abundancia, presión de caza y el impacto de la caza en mamíferos que habitan los bosques de la cuenca del río Alto Itaya. La información procede de censos por transectos y registros de caza llevados a cabo en seis comunidades. Entre los mamíferos de caza, el choro (Lagothrix poeppigii Schinz) fue el más abundante con 15,4 individuos/km², mientras que el mono aullador (Alouatta seniculus Linnaeus) y el venado colorado (Mazama americana Erxleben) fuero...

  5. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    Energy Technology Data Exchange (ETDEWEB)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I., E-mail: paul.hungler@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2014-07-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  6. Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

    International Nuclear Information System (INIS)

    Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I.

    2014-01-01

    In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

  7. Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

    In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

  8. 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

  9. Privatización del agua y racismo ambiental en ciudades segregadas. La empresa Aguas del Illimani en las ciudades de La Paz y El Alto (1997-2005

    Directory of Open Access Journals (Sweden)

    Crespo Flores, Carlos O.

    2009-12-01

    Full Text Available Within the framework of environmental racism in the water sector, the water and sani - tation concession contract in the cities of La Paz and El Alto (Bolivia to Aguas del Illimani company, filial of Suez Group. Both cities are characterised as racially segregated spaces. The differentiated inclusion service policy of the company to neighbourhoods in La Paz and El Alto, particularly costs of connexion and tariffs, the exclusion of poor aymara zones and the environmental risks and impacts.

    Desde el concepto de racismo ambiental aplicado al sector agua, se analiza la concesión del servicio de agua y saneamiento en las ciudades de La Paz y El Alto a la empresa Aguas del Illimani (AISA, filial de la compañía francesa Suez. Ambas ciudades se caracterizan por ser socioeconómica y racialmente segregadas. Se muestra la política de inclusión diferenciada de la empresa respecto a las laderas de la ciudad de La Paz y El Alto, particularmente en los costos de conexión y tarifas del servicio, la exclusión de zonas pobres de aymaras migrantes, los impactos y riesgos ambientales.

  10. 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

  11. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  12. Development of a new surface ion-source and ion guide in the ALTO project

    International Nuclear Information System (INIS)

    Cuong, P.V.

    2009-12-01

    The present work is dedicated to the ALTO project which is the production of neutron-rich gallium isotopes by the ISOL thick-target technique using photo-fission and a surface ion source. We aim at the study of the structure of 82 Ge, 83 Ge, 84 Ge via the β decay of 82 Ga, 83 Ga, and 84 Ga. We focus on the development of a new surface ion source made from materials with a high work function φ which can give high ionisation efficiencies for elements with low ionisation potentials, like alkaline as well as gallium and indium. Tungsten, rhenium and iridium are considered as good candidates for a surface ionizer because the Saha-Langmuir equation indicates high surface ionisation efficiencies for these materials. This has motivated us to equip the surface ion source at ALTO with rhenium and iridium-coated rhenium ionizer tubes of the same dimensions as the surface ion source at ISOLDE. We performed a test experiment to measure the ionisation efficiency for gallium. We also built a simulation code for the ionisation efficiency of the different surface ionisation sources (different materials and dimensions). On the other hand, for future nuclear structure studies of refractory elements such as cobalt or nickel, the ISOL technique with a thick target is no longer suitable. Indeed, the high melting point of these elements makes it difficult to volatilize and release them from a thick target. For such a situation, a technique based on thin targets is needed and the laser ion guide based on a gas cell to slow down, neutralize and stop the recoiling nuclear reaction products combined with a laser beam to re-ionize them selectively, seems a good choice. A code based on the Geant-4 tool-kit has been built to simulate the ionisation of the buffer gas. In this work, we also briefly show the results of the photo-fission yield measurements at ALTO. The fission fragments were ionized in a hot plasma ion source, mass separated and detected by germanium and scintillator detectors

  13. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  14. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

    1976-09-01

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  15. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

  16. Physical measurements at the RA reactor related to VISA-2, e. Measurements of flux and reactivity during RA reactor operation and exploitation; Fizicka merenja na reaktoru RA u vezi projekta VISA-2, e. Pracenje fluksa i reaktivnosti u toku eksploatacije reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-05-15

    This report includes the following: characteristics of neutron flux in vertical experimental channels of the RA reactor; characteristics of neutron flux in VISA-2 channels; reactivity changes in the reactor during VISA-2 irradiation including calibration of control rods.

  17. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  18. A study of UO2 wafer fuel for very high-power research reactors

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

  19. Microstructure in Zircaloy Creep Tested in the R2 Reactor

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2004-12-01

    Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

  20. In-situ stripping of H{sub 2}S in gasoil hydrodesulphurization - reactor design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Nava, J.A.O.; Krishna, R. [Amsterdam Univ., Dept. of Chemical Engineering, Amsterdam (Netherlands)

    2004-02-01

    In order to meet future diesel specifications the sulphur content of diesel would need to be reduced to below 50 ppm. This requirement would require improved reactor configurations. In this study we examine the benefits of counter-current contacting of gas oil with H{sub 2}, over conventional co-current contacting in a trickle bed hydrodesulphurization (HDS) reactor. In counter-current contacting, we achieve in-situ stripping of H{sub 2}S from the liquid phase; this is beneficial to the HDS kinetics. A comparison simulation study shows that counter-current contacting would require about 20% lower catalyst load than co-current contacting. However, counter-current contacting of gas and liquid phases in conventionally used HDS catalysts, of 1.5 mm sizes, is not possible due to flooding limitations. The catalysts need to be housed in special wire gauze envelopes as in the catalytic bales or KATAPAK-S configurations. A preliminary hardware design of a counter-current HDS reactor using catalytic bales was carried out in order to determine the technical feasibility. Using a realistic sulphur containing feedstock, the target of 50 ppm S content of desulphurized oil could be met in a reactor of reasonable dimensions. The study also underlines the need for accurate modelling of thermal effects during desulphurization. Our study also shows that interphase mass transfer is unlikely to be a limiting factor and there is a need to develop improved reactor configurations allowing for increased catalyst loading, at the expense of gas-liquid interfacial area. (Author)

  1. EL ALTO MAGDALENA- COLOMBIA DE LA MANO CON ENERGIAS ALTERNATIVAS

    Directory of Open Access Journals (Sweden)

    Barragán-Alturo, Ancízar

    2013-12-01

    Full Text Available El afán por destruir un paradigma, que mantiene encadenados a los habitantes de Girardot y la región, a una compañía de distribución de la energía eléctrica con sus altos precios para el kilowatt-hora, ha inspirado la investigación CUNDINAMARCA DE LA MANO DE LAS ENERGIAS ALTERNATIVAS, demostrando por diversos caminos que el montaje de paneles solares para generación de energía eléctrica en las cubiertas de las casas es la energía alternativa para la solución de diversos problemas, entre ellos: los costos elevados, las fluctuaciones de voltaje, los cortes de energía, los daños en los electrodomésticos

  2. Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

    International Nuclear Information System (INIS)

    El-Messiry, A.M.

    2000-01-01

    The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

  3. O uso de implantes orbitários de polietileno granulado de ultra-alto peso molecular no reparo de cavidades anoftálmicas

    Directory of Open Access Journals (Sweden)

    João Edward Soranz Filho

    2012-08-01

    Full Text Available OBJETIVO: Alterações oculares, em especial a perda de volume nas cavidades evisceradas, promovem uma série de modificações ao paciente tanto funcional do órgão quanto psicológica e estética. Para tanto a procura de um material de baixo custo e com biocompatibilidade tem sido uma constante na literatura. Portanto, esse trabalho teve como objetivo testar experimentalmente implante de polietileno granulado de ultra-alto peso molecular, material de baixo custo, em órbitas de coelhos submetidos à evisceração cirúrgica em vários tempos experimentais, onde foram avaliados aspectos macroscópicos e microscópicose de toxicidade sistêmica do material. MÉTODOS: Para esse estudo foram utilizados coelhos Oryctolaguscuniculus submetidos à evisceração do globo ocular direito e posteriormente implantados com esfera de polietileno granulado de ultra-alto peso molecular e analisados por 15, 30, 90 e 180 dias pós-implante, com parâmetros macro, microscópios e bioquímicos. Os animais controles foram submetidos ao mesmo procedimento sem, entretanto a colocação do implante. RESULTADOS: Os resultados desse trabalho mostram que o material utilizado no implante de cavidade não apresenta alteração significativa nos parâmetros de peso e bioquímicos quando comparados ao grupo controle. O material implantado apresentou uma grande interação com o tecido do hospedeiro. CONCLUSÃO: Os resultados indicam que implante de polietileno granulado de alto peso molecular desenvolvido por uma indústria nacional tem alto potencial para se realizar testes em humanos.

  4. Project requirements for reconstruction of the RA reactor ventilation system, Task 2.8. Measurement of radioactive iodine and other isotopes contents in the gas system of the RA reactor, Annex of the task

    International Nuclear Information System (INIS)

    Vujisic, Lj. et al

    1981-01-01

    This report is a supplement to the task 2.8. When planning and constructing the ventilation system, it was found that it is necessary to perform additional experiments during RA reactor operation at 2 MW power level for a longer period. In addition to the helium system, the potential source of radioactive pollutants is the space below the upper water shielding of the reactor. All the experimental and fuel channels are ending in this space. During repair and fuel exchange radioactivity can be released in this space. For that reason this space is important when planing and designing the filtration system for incidental conditions or planned dehermetisation of the reactor. The third point where radioactive isotope identification was done, was the entrance into the chimney during steady state operation and planned dehermetisation of the reactor. The following samples were measured: gas system during reactor operation at 2 MW power; entrance into the chimney during last 48 hours of reactor operation at 2 MW power; sample on the platform under the upper water shield with the opened fuel channel after the reactor shutdown; and simultaneously with the latter, measurement at the entrance to the chimney. This annex contains the list of identified radioactive isotopes, volatile and gaseous as well as concentration of volatile 131 I on the adsorbents [sr

  5. 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.)

  6. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K J; Baik, S T; Chung, U S; Jung, K H; Park, S K; Lee, B J; Kim, J K; Yang, S H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

  7. Crecimiento del maíz en vertisoles con alto aluminio en la Baixada Maranhense pre-Amazonia, Brasil.

    Directory of Open Access Journals (Sweden)

    Alessandro Costa-da Silva

    2014-07-01

    Full Text Available El objetivo de este trabajo fue evaluar el crecimiento del maíz en suelos con alto contenido de aluminio. Se midió el efecto del Al3+ en raíces y la cantidad de materia seca (raíz, hoja y tallo de maíz. Se efectuó la caracterización físico-química de cuatro muestras de suelo con alto aluminio colectadas del horizonte Ap, en tres municipios de la región conocida como Baixada Maranhense (Pre-Amazonia, Brasil: Santa Rita (SR, Arari (AR y Vitoria do Mearim (VM y un testigo colectado en el municipio de São Luís, Área del Núcleo de Tecnología Rural (T. El estudio, ejecutado en 2009, se llevó a cabo en invernadero y se utilizó 2 dm3 de suelo por maceta. Asimismo las muestras fueron divididas en muestras con y sin fertilización. La variación en la longitud de la raíz y de materia seca de las hojas difirió significativamente entre tratados con y sin fertilizante, excepto en la muestra de la localidad T. La producción de materia seca de raíz, tallo y hoja fue mayor en todos los suelos cuando se fertilizó. El suelo testigo también superó a todos los demás en cuanto a producción de materia seca en la raíz, posiblemente como resultado de una menor cantidad de Al3+ (1,2 cmolc/dm3 en comparación con los suelos SR, AR y VM (6,8; 8,0 y 7,0 cmolc/dm3 respectivamente. Se concluye la fertilización reduce el efecto detrimental del aluminio en la producción de maíz en la Baixada Maranhense.

  8. Tratamento bem-sucedido da síndrome cardiopulmonar por hantavírus com uso de hemofiltração de alto volume

    Directory of Open Access Journals (Sweden)

    Guillermo Bugedo

    Full Text Available RESUMO A síndrome cardiopulmonar por hantavírus tem elevada taxa de mortalidade. Sugere-se que uma conexão precoce com oxigenação por membrana extracorpórea melhore os resultados. Relatamos o caso de uma paciente que apresentou síndrome cardiopulmonar por hantavírus e choque refratário, que preenchia os critérios para oxigenação por membrana extracorpórea e que teve resposta satisfatória com uso de hemofiltração contínua de alto volume. A implantação de hemofiltração contínua de alto volume, juntamente da ventilação protetora, reverteu o choque dentro de poucas horas e pode ter levado à recuperação. Em pacientes com síndrome cardiopulmonar por hantavírus, um curso rápido de hemofiltração contínua de alto volume pode ajudar a diferenciar pacientes que podem ser tratados com cuidados convencionais da unidade de terapia intensiva dos que necessitarão de terapias mais complexas, como oxigenação por membrana extracorpórea.

  9. Study on effects of development of reactor constant in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

    Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

  10. Thermal and stress analyses of the reactor pressure vessel lower head of the Three Mile Island Unit 2

    International Nuclear Information System (INIS)

    Hashimoto, K.; Onizawa, K.; Kurihara, R.; Kawasaki, S.; Soda, K.

    1992-01-01

    Thermal and stress analyses were performed using the finite element analysis code ABAQUS to clarify the factors which caused tears in the stainless steel liner of the reactor pressure vessel lower head of the Three Mile Island Unit 2 (TMI-2) reactor pressure vessel during the accident on 28 March 1979. The present analyses covered the events which occurred after approximately 20 tons of molten core material were relocated to the lower head of the reactor pressure vessel. They showed that the tensile stress was highest in the case where the relocated core material consisting of homogeneous UO 2 debris was assumed to attack the lower head and the debris was then quenched. The peak tensile stress was in the vicinity of the welded zone of the penetration nozzle. This result agrees with the findings from the examination of the TMI-2 reactor pressure vessel that major tears in the stainless steel liner were observed around two penetration nozzles of the lower head. (author)

  11. Efecto de empaquetamiento de las partículas en la durabilidad de los hormigones de alto desempeño

    Directory of Open Access Journals (Sweden)

    A. Castro

    Full Text Available Puesto que las estructuras de hormigón están expuestas a un alto grado de agresividad, se han realizado muchas investigaciones al respecto, en el área de la tecnología del hormigón. Los investigadores se han enfocado en la búsqueda de materiales que posean un mejor comportamiento mecánico así como una mayor durabilidad, para lo cual es necesario un hormigón obtenido mediante la ingeniería de la microestructura de los materiales. Como los materiales compuestos están formados por partículas de granulometría fina y una baja razón agua/aglomerante logran una matriz densa cuando se optimiza el empacamiento de sus partículas. Basándose en el concepto de empacamiento de las partículas en el diseño de los hormigones de alto desempeño, este trabajo analiza la durabilidad de esos materiales a través de la determinación de la penetración del ion cloruro y absorción de agua por inmersión y capilaridad. Se compararon las propiedades del hormigón diseñado bajo este concepto con las de los hormigones de alto desempeño diseñados de acuerdo a la metodología convencional, mostrando mejores propiedades en términos de durabilidad.

  12. The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  13. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    International Nuclear Information System (INIS)

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  14. Tratamento bem-sucedido da síndrome cardiopulmonar por hantavírus com uso de hemofiltração de alto volume

    OpenAIRE

    Bugedo, Guillermo; Florez, Jorge; Ferres, Marcela; Roessler, Eric; Bruhn, Alejandro

    2016-01-01

    RESUMO A síndrome cardiopulmonar por hantavírus tem elevada taxa de mortalidade. Sugere-se que uma conexão precoce com oxigenação por membrana extracorpórea melhore os resultados. Relatamos o caso de uma paciente que apresentou síndrome cardiopulmonar por hantavírus e choque refratário, que preenchia os critérios para oxigenação por membrana extracorpórea e que teve resposta satisfatória com uso de hemofiltração contínua de alto volume. A implantação de hemofiltração contínua de alto volume, ...

  15. The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

    International Nuclear Information System (INIS)

    Atkinson, C.A.; Collins, P.J.

    1987-01-01

    Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

  16. TMI-2 reactor vessel and balance of plant status

    International Nuclear Information System (INIS)

    Kuehn, G.A.

    1990-01-01

    In the fall of 1988 a corporate decision was made which concentrated effort on support of reactor vessel defueling and minimized activity in balance-of-plant areas. The auxiliary and fuel handling building are in a safe/stable state but final preparations for monitored storage won't be pursued until defueling and fuel shipping are complete. In addition to dispositioning fuel, the project is actively preparing for disposal of the Accident Generated Water (2.3 million gallons) by evaporation

  17. Abatement of fluorinated compounds using a 2.45 GHz microwave plasma torch with a reverse vortex plasma reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.H.; Cho, C.H.; Shin, D.H. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Hong, Y.C., E-mail: ychong@nfri.re.kr [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); Shin, Y.W. [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Oxikdo-dong, Gunsan-city, Jeollabuk-do (Korea, Republic of); School of Advanced Green Energy and Environments, Handong Global University, Heunghae-eup, Buk-gu, Pohang-city, Gyeongbuk (Korea, Republic of)

    2015-08-30

    Highlights: • We developed a microwave plasma torch with reverse vortex reactor (RVR). • We calculated a volume fraction and temperature distribution of discharge gas and waste. • The performance of reverse vortex reactor increased from 29% to 43% than conventional vortex reactor. - Abstract: Abatement of fluorinated compounds (FCs) used in semiconductor and display industries has received an attention due to the increasingly stricter regulation on their emission. We have developed a 2.45 GHz microwave plasma torch with reverse vortex reactor (RVR). In order to design a reverse vortex plasma reactor, we calculated a volume fraction and temperature distribution of discharge gas and waste gas in RVR by ANSYS CFX of computational fluid dynamics (CFD) simulation code. Abatement experiments have been performed with respect to SF{sub 6}, NF{sub 3} by varying plasma power and N{sub 2} flow rates, and FCs concentration. Detailed experiments were conducted on the abatement of NF{sub 3} and SF{sub 6} in terms of destruction and removal efficiency (DRE) using Fourier transform infrared (FTIR). The DRE of 99.9% for NF{sub 3} was achieved without an additive gas at the N{sub 2} flow rate of 150 liter per minute (L/min) by applying a microwave power of 6 kW with RVR. Also, a DRE of SF{sub 6} was 99.99% at the N{sub 2} flow rate of 60 L/min using an applied microwave power of 6 kW. The performance of reverse vortex reactor increased about 43% of NF{sub 3} and 29% of SF{sub 6} abatements results definition by decomposition energy per liter more than conventional vortex reactor.

  18. 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)

  19. Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

    1978-12-01

    This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

  20. PARSJAD, il Parco Archeologico dell’Alto Adriatico: un luogo diffuso, unito dalla tecnologia

    Directory of Open Access Journals (Sweden)

    Emilia Buniotto

    2014-05-01

    Full Text Available Eight partners, coordinated by the Veneto Region, are working on an ambitious project with an innovative approach and tools, and with the ability to conserve content and exploit the immense archaeological heritage of the area concerned.This was the birth of the Parco Archeologico dell’Alto Adriatico (Upper Adriatic Archaeological Park, a project funded by the Italy‐Slovenia 2007-2013 Cooperation Programme, with the aim of tracing a unified and cross-border pathof knowledge running from the coast of Emilia to the one in Slovenia.

  1. Cronos 2: a neutronic simulation software for reactor core calculations; Cronos 2: un logiciel de simulation neutronique des coeurs de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lautard, J J; Magnaud, C; Moreau, F; Baudron, A M [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France)

    1999-07-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  2. TARMS, an on-line boiling water reactor operation management system. [3 D core simulator LOGOS 2

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-12-01

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy.

  3. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

    2012-01-01

    This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

  4. 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.

  5. 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.)

  6. 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

  7. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-09-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

  8. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  9. 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

  10. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1998-09-01

    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  11. 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.)

  12. 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.

  13. Comercio e intercambio en la Hispania romana (Alto imperio

    Directory of Open Access Journals (Sweden)

    Genaro CHIC GARCÍA

    2010-02-01

    Full Text Available RESUMEN: Se muestra al Alto Imperio como una etapa de desarrollo del individualismo en relación con la formación de mercados impersonales dominados por la racionalidad económica, manifiesta en signos tales como la moneda y la escritura fonético- vocálica, que va desplazando a los sistemas de relación tradicionales basados en el prestigio social y los intercambios personalizados. La formación de un Estado burocrático centralizado, sin embargo, hará colapsar el incipiente liberalismo comercial.ABSTRACT: The Early Empire is shown as a stage of development of the individualism in relation to the formation of impersonal markets dominated by the economic rationality, wich becomes evident in signs such as the coin or the phonetic-vocalic writing, that is displacing the traditional systems of relation based on the social prestige and the personalized exchange. Nevertheless, the formation of a centralized bureaucratic state will cause the incipient commercial liberalism to collapse.

  14. Catalytic combustion of the retentate gas from a CO2/H2 separation membrane reactor for further CO2 enrichment and energy recovery

    International Nuclear Information System (INIS)

    Hwang, Kyung-Ran; Park, Jin-Woo; Lee, Sung-Wook; Hong, Sungkook; Lee, Chun-Boo; Oh, Duck-Kyu; Jin, Min-Ho; Lee, Dong-Wook; Park, Jong-Soo

    2015-01-01

    The CCR (catalytic combustion reaction) of the retentate gas, consisting of 90% CO 2 and 10% H 2 obtained from a CO 2 /H 2 separation membrane reactor, was investigated using a porous Ni metal catalyst in order to recover energy and further enrich CO 2 . A disc-shaped porous Ni metal catalyst, namely Al[0.1]/Ni, was prepared by a simple method and a compact MCR (micro-channel reactor) equipped with a catalyst plate was designed for the CCR. CO 2 and H 2 concentrations of 98.68% and 0.46%, respectively, were achieved at an operating temperature of 400 °C, GHSV (gas-hourly space velocity) of 50,000 h −1 and a H 2 /O 2 ratio (R/O) of 2 in the unit module. In the case of the MCR, a sheet of the Ni metal catalyst was easily installed along with the other metal plates and the concentration of CO 2 in the retentate gas increased up to 96.7%. The differences in temperatures measured before and after the CCR were 31 °C at the product outlet and 19 °C at the N 2 outlet in the MCR. The disc-shaped porous metal catalyst and MCR configuration used in this study exhibit potential advantages, such as high thermal transfer resulting in improved energy recovery rate, simple catalyst preparation, and easy installation of the catalyst in the MCR. - Highlights: • The catalytic combustion of a retentate gas obtained from the H 2 /CO 2 separation membrane. • A disc-shaped porous nickel metal catalyst and a micro-channel reactor for catalytic hydrogen combustion. • CO 2 enrichment up to 98.68% at 400 °C, 50,000 h −1 and H 2 /O 2 ratio of 2.

  15. 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.)

  16. Multidrug-resistant pulmonary tuberculosis in Los Altos, Selva and Norte regions, Chiapas, Mexico.

    Science.gov (United States)

    Sánchez-Pérez, H J; Díaz-Vázquez, A; Nájera-Ortiz, J C; Balandrano, S; Martín-Mateo, M

    2010-01-01

    To analyse the proportion of multidrug-resistant tuberculosis (MDR-TB) in cultures performed during the period 2000-2002 in Los Altos, Selva and Norte regions, Chiapas, Mexico, and to analyse MDR-TB in terms of clinical and sociodemographic indicators. Cross-sectional study of patients with pulmonary tuberculosis (PTB) from the above regions. Drug susceptibility testing results from two research projects were analysed, as were those of routine sputum samples sent in by health personnel for processing (n = 114). MDR-TB was analysed in terms of the various variables of interest using bivariate tests of association and logistic regression. The proportion of primary MDR-TB was 4.6% (2 of 43), that of secondary MDR-TB was 29.2% (7/24), while among those whose history of treatment was unknown the proportion was 14.3% (3/21). According to the logistic regression model, the variables most highly associated with MDR-TB were as follows: having received anti-tuberculosis treatment previously, cough of >3 years' duration and not being indigenous. The high proportion of MDR cases found in the regions studied shows that it is necessary to significantly improve the control and surveillance of PTB.

  17. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1995-10-01

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  18. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  19. Model study of an automatic controller of the IBR-2 pulsed reactor

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Popov, A.K.

    2007-01-01

    For calculation of power transients in the IBR-2 reactor a special mathematical model of dynamics taking into account the discontinuous jump of reactivity by an automatic controller with the step motor is created. In the model the nonlinear dependence of the energy of power pulse on the reactivity and the influence of warming up of the reactor on the reactivity by means of introduction of a nonlinear feedback 'power-pulse energy - reactivity' are taken into account. With the help of the model the transients of relative deviation of power-pulse energy are calculated at various (random, mixed and regular) reactivity disturbances at the reactor mean power 1.475 MW. It is shown that to improve the quality of processes the choice of such regular values of parameters of the automatic controller is expedient, at which the least effect of smoothing of a signal acting on an automatic controller and the least speed of an automatic controller are provided, and the reduction of efficiency of one step of the automatic controller and introduction of a five-percent dead space are also expedient

  20. Backfitting of the FRG reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

    1990-05-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

  1. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  2. Compact stellarators as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

    2001-01-01

    Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

  3. Prevalência e variáveis associadas à inatividade física em indivíduos de alto e baixo nível socioeconômico Prevalencia y variables asociadas a la inactividad física en individuos de alto y bajo nivel socioeconómico Prevalence and variables associated with physical inactivity in individuals with high and low socioeconomic status

    Directory of Open Access Journals (Sweden)

    Helena França Correia dos Reis

    2009-03-01

    Full Text Available FUNDAMENTO: Estudos que consideram apenas a atividade física de lazer encontraram que a inatividade física é maior entre os indivíduos com menor renda. Existe a possibilidade de que, ao se considerar as atividades de transporte, trabalho e domésticas ocorra modificações nessa associação. OBJETIVO:Determinar se há diferença entre as prevalências de inatividade física entre indivíduos de alto e baixo nível socioeconômico. MÉTODOS: A amostra foi constituída por indivíduos de ambos os sexos, com 18 anos ou mais, de dois grupos de diferentes níveis socioeconômicos. O Grupo de baixo nível socioeconômico (BNSE foi composto por pais de alunos de uma escola pública. Os indivíduos de alto nível socioeconômico (ANSE foram pais de uma escola de nível superior privada. Para determinação do nível de atividade física foi utilizado o Questionário Internacional de Atividade Física (IPAQ. RESULTADOS: Noventa e um indivíduos foram avaliados no grupo de BNSE e 59 no ANSE. No grupo de baixo NSE, 42,9% (39 dos indivíduos foram classificados como insuficientemente ativos, comparados a 57,6% (34 nos indivíduos de alto NSE. Tomando-se como parâmetro de inatividade física um tempo de atividade física semanal menor que 150 minutos houve redução da classificação de inatividade em ambos os grupos, porém com manutenção de maior inatividade nos indivíduos de alto NSE (49,2% vs 28,6%; p= 0,01. CONCLUSÃO:Os indivíduos de alto nível socioeconômico são mais sedentários que os indivíduos de baixo nível socioeconômico.FUNDAMENTO: Estudios que incluyen sólo la actividad física de ocio revelaron que la inactividad física es mayor entre los individuos con menor renta. Existe la posibilidad de que, al tomarse en consideración las actividades de transporte, trabajo y domésticas ocurra cambios en esa asociación. OBJETIVO:Determinar si hay diferencia en las prevalencias de inactividad física entre individuos de alto y bajo

  4. Fragilidade ambiental na bacia hidrográfica do Alto Parnaíba

    OpenAIRE

    Alves de Melo, Nivaneide

    2007-01-01

    Esta pesquisa desenvolveu estudos integrados sobre a bacia hidrográfica do Alto Parnaíba, no Piauí, considerando a atuação das atividades humanas sobre o ambiente natural, a fim de determinar o grau de alteração desse ambiente, a partir da presença antrópica neste local, além de propor ações para restabelecimento de uma situação de equilíbrio ambiental. Pois, quando os eventos sobre a paisagem são de origem antrópica e de orientação econômica, os impactos poderão causar dano...

  5. Vazões máximas e mínimas para bacias hidrográficas da região alto Rio Grande, MG Maximum and minimum discharges for Alto Rio Grande region basins, Minas Gerais state, Brazil

    Directory of Open Access Journals (Sweden)

    Carlos Rogério de Mello

    2010-04-01

    Full Text Available Vazões máximas são grandezas hidrológicas aplicadas a projetos de obras hidráulicas e vazões mínimas são utilizadas para a avaliação das disponibilidades hídricas em bacias hidrográficas e comportamento do escoamento subterrâneo. Neste estudo, objetivou-se à construção de intervalos de confiança estatísticos para vazões máximas e mínimas diárias anuais e sua relação com as características fisiográficas das 6 maiores bacias hidrográficas da região Alto Rio Grande à montante da represa da UHE-Camargos/CEMIG. As distribuições de probabilidades Gumbel e Gama foram aplicadas, respectivamente, para séries históricas de vazões máximas e mínimas, utilizando os estimadores de Máxima Verossimilhança. Os intervalos de confiança constituem-se em uma importante ferramenta para o melhor entendimento e estimativa das vazões, sendo influenciado pelas características geológicas das bacias. Com base nos mesmos, verificou-se que a região Alto Rio Grande possui duas áreas distintas: a primeira, abrangendo as bacias Aiuruoca, Carvalhos e Bom Jardim, que apresentaram as maiores vazões máximas e mínimas, significando potencialidade para cheias mais significativas e maiores disponibilidades hídricas; a segunda, associada às bacias F. Laranjeiras, Madre de Deus e Andrelândia, que apresentaram as menores disponibilidades hídricas.Maximum discharges are applied to hydraulic structure design and minimum discharges are used to characterize water availability in hydrographic basins and subterranean flow. This study is aimed at estimating the confidence statistical intervals for maximum and minimum annual discharges and their relationship wih the physical characteristics of basins in the Alto Rio Grande Region, State of Minas Gerais. The study was developed for the six (6 greatest Alto Rio Grande Region basins at upstream of the UHE-Camargos/CEMIG reservoir. Gumbel and Gama probability distribution models were applied to the

  6. Design options for a bunsen reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Robert Charles

    2013-10-01

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  7. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density LEU fuels that are being developed by the RERTR program. High-density LEU dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits

  8. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density Leu fuels that are being developed by the Rarita program. High-density Leu dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits. (author)

  9. Performance analysis of photocatalytic CO2 reduction in optical fiber monolith reactor with multiple inverse lights

    International Nuclear Information System (INIS)

    Yuan, Kai; Yang, Lijun; Du, Xiaoze; Yang, Yongping

    2014-01-01

    Highlights: • A new optical fiber monolith reactor model for CO 2 reduction was developed. • Methanol concentration versus fiber location and operation parameters was obtained. • Reaction efficiency increases by 31.1% due to the four fibers and inverse layout. • With increasing space of fiber and channel center, methanol concentration increases. • Methanol concentration increases as the vapor ratio and light intensity increase. - Abstract: Photocatalytic CO 2 reduction seems potential to mitigate greenhouse gas emissions and produce renewable energy. A new model of photocatalytic CO 2 reduction in optical fiber monolith reactor with multiple inverse lights was developed in this study to improve the conversion of CO 2 to CH 3 OH. The new light distribution equation was derived, by which the light distribution was modeled and analyzed. The variations of CH 3 OH concentration with the fiber location and operation parameters were obtained by means of numerical simulation. The results show that the outlet CH 3 OH concentration is 31.1% higher than the previous model, which is attributed to the four fibers and inverse layout. With the increase of the distance between the fiber and the monolith center, the average CH 3 OH concentration increases. The average CH 3 OH concentration also rises as the light input and water vapor percentage increase, but declines with increasing the inlet velocity. The maximum conversion rate and quantum efficiency in the model are 0.235 μmol g −1 h −1 and 0.0177% respectively, both higher than previous internally illuminated monolith reactor (0.16 μmol g −1 h −1 and 0.012%). The optical fiber monolith reactor layout with multiple inverse lights is recommended in the design of photocatalytic reactor of CO 2 reduction

  10. Feminicidios en la frontera chilena: el caso de Alto Hospicio

    Directory of Open Access Journals (Sweden)

    Ainhoa Vásquez Mejías

    2016-01-01

    Full Text Available Entre 1999 y el 2001, varias adolescentes desaparecieron en el norte de Chile y, posteriormente, fueron encontradas muertas. La novela Alto Hospicio (2008, de Rodrigo Ramos Bañados, recrea estos feminicidios desde la visión del cómplice del único inculpado por estos crímenes. En el presente artículo se plantea que el escritor utiliza la noción de frontera como aparato crítico, con el fin de desestabilizar barreras espaciales, simbólicas y textuales que se agrupan en cuatro ejes: Bolivia-Chile, santas-putas, racionalidad-locura, ficción-realidad. Así, la novela cuestiona y deconstruye los límites para hablarnos de estos feminicidios desde un punto indeterminado, que funciona como espejo de la incertidumbre e irresolución de este caso en la justicia chilena.

  11. MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

  12. Fast reactors worldwide

    International Nuclear Information System (INIS)

    Hall, R.S.; Vignon, D.

    1985-01-01

    The paper concerns the evolution of fast reactors over the past 30 years, and their present status. Fast reactor development in different countries is described, and the present position, with emphasis on cost reduction and collaboration, is examined. The French development of the fast breeder type reactor is reviewed, and includes: the acquisition of technical skills, the search for competitive costs and the spx2 project, and more advanced designs. Future prospects are also discussed. (U.K.)

  13. Application of stable adaptive schemes to nuclear reactor systems, (2)

    International Nuclear Information System (INIS)

    Kukuda, Toshio

    1979-01-01

    The parameter identification and adaptive control schemes applied in a previous study to a nonlinear point reactor are extended to the case of a loosely-coupled-core reactor with internal feedbacks, constituting a nonlinear overall system. Both schemes are shown to be stable, with the system newly represented on the pattern of the Model Reference Adaptive System (MRAS) with use made of the Lyapunov's method. For either parameter identification or adaptive control of a loosely-coupled-core reactor, there exists no canonical form of multiple input-multiple output system which can be directly applied for deriving the MRAS with the matrix version of the Kalman-Yakubovich lemma as it was in the case of the point reactor. This difficulty is circumvented by the practical assumption that the neutron density can be directly measured on each core as reactivity change is applied as input into the coupled core as a whole. For parameter identification, the model parameters are adaptively adjusted to those of each core, while for the adaptive control, plant parameters of each core can be adaptively compensated, again through control inputs, to asymptotically reduce the output error between the model and the plant. The point reactor is shown to correspond to a special case. (author)

  14. Propiedades dieléctricas de láminas delgadas de (Pb,CaTiO3 con alto contenido de Ca

    Directory of Open Access Journals (Sweden)

    Calzada, M. L.

    2004-04-01

    Full Text Available Lead titanate thin films modified with calcium, Pb1-xCaxTiO3, with x>35%, have been deposited onto Pt/TiO2/SiO2/(100Si y Pt/(100MgO substrates, by the sol-gel method and a rapid thermal processing (RTP. Properties similar to the counterpart bulk ceramics are planned, such as the decrease of the phase transition temperature and the increase of permittivity at room temperature. Films with Ca content close to 40 % and 50% exhibit at room temperature, high enough permittivity with reduced temperature dependence and a C-V behaviour that make them of interest for DRAM and high frequency devices.Se han preparado láminas delgadas de titanato de plomo modificado con calcio, Pb1-xCaxTiO3, (PTCa con x>35%, depositadas sobre substratos de Pt/TiO2/SiO2/(100Si y Pt/(100MgO, empleando la técnica de sol-gel y un horno rápido de procesado (RTP. Se intenta obtener propiedades cercanas al material cerámico masivo, como son la bajada de la temperatura de la transición ferro-paraeléctrica y el aumento de la permitividad, ε´ a temperatura ambiente. Así, se consiguen láminas de contenido de Ca próximas al 40% y 50% con valores altos de permitividad, poco dependientes de la temperatura, que presentan un alto carácter difuso en la transición para temperaturas próxima a ambiente y un comportamiento de la capacidad con el voltaje, C-V,que las hacen de interés en la fabricación de DRAM y dispositivos para alta frecuencia.

  15. 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)

  16. Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

    International Nuclear Information System (INIS)

    Ponsard, B.

    2000-01-01

    The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

  17. Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

    International Nuclear Information System (INIS)

    Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

    2013-01-01

    Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

  18. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2016-09-15

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  19. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    International Nuclear Information System (INIS)

    Cuevas V, G.; Banfield, J.; Avila N, A.

    2016-09-01

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  20. Aplicabilidade do Brums: estados de humor em atletas de voleibol e tênis no alto rendimento

    Directory of Open Access Journals (Sweden)

    Tatiana Marcela Rotta

    2014-12-01

    Full Text Available Introdução: Os estados de humor são indicadores que auxiliam sobremaneira o rendimento e a prevenção da saúde do atleta. Objetivo: Analisar a aplicabilidade do instrumento BRUMS na avaliação do perfil de estados de humor em atletas de alto rendimento do sexo masculino, de voleibol n=59 e tênis n=69. Métodos: comparar as variáveis independentes: modalidade esportiva (voleibol e tênis; tempo de prática no alto rendimento (até 2 anos; mais de 2 anos e categorias de idade (jovens e adultos com as variáveis dependente do perfil de humor (tensão, depressão, raiva, vigor, fadiga e confusão mental. O estudo causal-comparativo, utilizou, para coleta de dados, o instrumento BRUMS, validado no Brasil, com participação das autoras desse estudo. Resultados: Ao executar a MANOVA, foram verificadas diferenças entre as modalidades (F=4,289/ p=0,001; Hotellings's Trace = 0,216 e tempo de prática (F=5,845/ p<0,001; Hotelling's Trace = 0,295 no vigor. A modalidade versus tempo de prática apresentou significância na interação das duas variáveis (p=0,003, em torno de 7% da variância. Também na variável tensão p=0,05 (voleibol vs. tempo de prática, na variável raiva p=0,001 (voleibol vs. categoria de idade. No tênis, a depressão p=0,001, raiva p=0,04 e confusão mental p< 0,02 com médias maiores em adultos e mais experientes. Conclusão: A modalidade foi responsável por 11% da alteração no perfil de humor (p=0,001.

  1. Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

    International Nuclear Information System (INIS)

    González Mantecón, Javier

    2015-01-01

    The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

  2. CO2 Energy Reactor - Integrated Mineral Carbonation: Perspectives on Lab-Scale Investigation and Products Valorization

    OpenAIRE

    Rafael M Santos; Pol CM Knops; Keesjan L Rijnsburger; Yi Wai eChiang

    2016-01-01

    To overcome the challenges of mineral CO2 sequestration, Innovation Concepts B.V. is developing a unique proprietary gravity pressure vessel (GPV) reactor technology and has focussed on generating reaction products of high economic value. The GPV provides intense process conditions through hydrostatic pressurization and heat exchange integration that harvests exothermic reaction energy, thereby reducing energy demand of conventional reactor designs, in addition to offering other benefits. In ...

  3. Estado nutricional de crianças índias do Alto Xingu em 1980 e 1992 e evolução pondero-estatural entre o primeiro e o quarto anos de vida Nutritional status of indigenous children from the Alto Xingu in 1980 and 1992 and follow-up of weight and height from the first through the fourth years of life

    Directory of Open Access Journals (Sweden)

    Mauro Batista de Morais

    2003-04-01

    Full Text Available Os objetivos deste estudo realizado com a população infantil do Alto Xingu foram: (1 analisar a evolução do peso e da estatura entre o primeiro e o quarto anos de vida, (2 comparar o estado nutricional em 1980 e 1992. Avaliaram-se o peso e a estatura de: (1 81 crianças no primeiro e no quarto ano de vida; (2 264 crianças avaliadas em 1980 e de 172 em 1992 (idade This study focused on the under-five population of the Alto Xingu region in Brazil, with the following objectives: (1 to evaluate height and weight increment from the first through the fourth years of life and (2 to compare nutritional status in 1980 and 1992. Height and weight increases were evaluated in 81 children. Weight and height were measured in 264 children evaluated in 1980 and in 172 in 1992 (< 10 years of age. Median Z-scores in the first and fourth years of life, respectively, showed: (1 a decrease in weight-for-age, (-0.12 in the first year and -0.51 in the fourth year of life; p = 0.002; (2 a decrease in weight-for-height (+1.31 and +0.08; p < 0.001; (3 an increase in height-for-age (-1.50 and -0.94; p < 0.001. Median Z-scores in 1980 and 1992 showed: (1 no change in weight-for-age (-0.61 in 1980 and -0.62 in 1992; p = 0.90; (2 no change in weight-for-height (+0.27 and +0.34; p = 0.10; and (3 a decrease in height-for-age (-1.04 and -1.22; p = 0.02. Height-for-age increased and weight-for-height decreased between the first and fourth years of life. A decrease in height-for-age was observed from 1980 to 1992, demonstrating the importance of nutritional surveillance among the population of the Alto Xingu.

  4. What occurred in the reactors

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko

    2013-01-01

    Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

  5. 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.)

  6. Catalytic wet oxidation of phenol in a trickle bed reactor over a Pt/TiO2 catalyst.

    Science.gov (United States)

    Maugans, Clayton B; Akgerman, Aydin

    2003-01-01

    Catalytic wet oxidation of phenol was studied in a batch and a trickle bed reactor using 4.45% Pt/TiO2 catalyst in the temperature range 150-205 degrees C. Kinetic data were obtained from batch reactor studies and used to model the reaction kinetics for phenol disappearance and for total organic carbon disappearance. Trickle bed experiments were then performed to generate data from a heterogeneous flow reactor. Catalyst deactivation was observed in the trickle bed reactor, although the exact cause was not determined. Deactivation was observed to linearly increase with the cumulative amount of phenol that had passed over the catalyst bed. Trickle bed reactor modeling was performed using a three-phase heterogeneous model. Model parameters were determined from literature correlations, batch derived kinetic data, and trickle bed derived catalyst deactivation data. The model equations were solved using orthogonal collocations on finite elements. Trickle bed performance was successfully predicted using the batch derived kinetic model and the three-phase reactor model. Thus, using the kinetics determined from limited data in the batch mode, it is possible to predict continuous flow multiphase reactor performance.

  7. Reactividad y propiedades mecánicas de escoria de alto horno activada por álcalis

    Directory of Open Access Journals (Sweden)

    Escalante-García, J. I.

    2002-10-01

    Full Text Available Investigations were carried out to characterize the reactivity and mechanical properties in blast furnace slag pastes activated by 5 alkali systems such as NaOH, waterglass and combinations of Na2CO3, Na2SO4 y Ca(OH2 in concentrations of 5%. Cubes of 5cm were prepared and cured under water for up to 120 days at 20 and 60°C. The highest compressive strengths were noted for the waterglass activation, followed by one of the combination of 3 compounds, while NaOH activation resulted in the lowest strengths. Increased temperatures favored the strength for the waterglass activation, outperforming those of the Portland cement paste used as control at all times, whereas it had a negative effect for all other activations. The reactivity of the systems, as obtained by measurements of non evaporable water and selective chemical dissolution, indicated that the slag is more reactive under the NaOH activation. Scanning electron microscopy by means of backscattered electrons confirmed the different reactivities and the mechanical properties observed.

    Se realizaron estudios de caracterización de reactividad y propiedades mecánicas en pastas de escoria de alto horno activada por 5 sistemas alcalinos como NaOH, vidrio soluble y combinaciones de Na2CO3, Na2SO4 y Ca(OH2 en concentraciones de 5%. Se prepararon cubos de 5cm por lado y se curaron durante 120 días a 20 y 60°C bajo agua. Los valores mas altos de resistencia mecánica fueron registrados por el sistema de escoria activada con vidrio soluble, seguido por una de las combinaciones de 3 reactivos químicos, mientras que el desarrollo de propiedades mecánicas fue más pobre con la activación por NaOH. El aumento en la temperatura de curado favoreció la resistencia a la compresión de escoria con vidrio soluble, superando a la del cemento Portland a todos los tiempos de curado; para los otros sistemas el incremento de la temperatura de curado resultó perjudicial. En contraste con las

  8. Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

    International Nuclear Information System (INIS)

    Bergdahl, B.G.

    1998-05-01

    The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was 'filtered', meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network

  9. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    International Nuclear Information System (INIS)

    Guigon, B.; Vacelet, H.; Dornbusch, D.

    2000-01-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

  10. TMI-2 [Three Mile Island Unit 2] reactor building dose reduction task force

    International Nuclear Information System (INIS)

    Daniels, R.S.

    1988-01-01

    In late October 1982, the director of Three Mile Island Unit 2 (TMI-2) created the dose reduction task force with the objective of identifying the principal radiological sources in the reactor building and recommending actions to minimize the dose to workers on labor-intensive projects. Members of the task force were drawn form various groups at TMI. Findings and recommendations were presented to the US Nuclear Regulatory Commission in a briefing on November 18, 1982. The task force developed a three-step approach toward dose reduction. Step 1 identified the radiological sources. Step 2 modeled the source and estimated its contribution to the general area dose rates. Step 3 recommended actions to achieve dose reductions consistent with general exposure rate goals

  11. A novel auto-thermal reforming membrane reactor for high purity H2

    International Nuclear Information System (INIS)

    Tony Boyd; Grace, J.R.; Lim, C.J.; Adris, A.M.

    2006-01-01

    A novel hydrogen reactor based on steam reforming of natural gas has been developed and tested. The reactor produces high purity hydrogen using in-situ perm-selective membranes installed in a fluidized catalyst bed, thus shifting the thermodynamic equilibrium of the SMR reaction and eliminating the need for downstream hydrogen purification. The reactor is particularly suited to auto-thermal reforming, where air is added to the reformer to provide the endothermic reaction heat, thus eliminating the need to indirectly heat the reactor. The gas flow pattern within the fluidized bed induces an internal circulation of catalyst particles between the central SMR reaction (permeation) zone and an outer annulus. The circulating hot catalyst particles from the oxidation zone carry the required endothermic heat of reaction for the reforming, while ensuring that the palladium membranes are not exposed to excessive temperatures or to oxygen. Another beneficial characteristic of the reactor is that very little of the nitrogen present in the oxidation air reaches the reaction zone, thus maintaining the hydrogen driving force for the perm-selective membranes. Pilot plant results carried out in a semi-industrial scale reactor will be presented. The reactor was operated up to 650 C and 14 bar. Pure hydrogen (99.999+%) was initially obtained from the reactor and an equilibrium shift was demonstrated. (authors)

  12. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

  13. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  14. Sentidos de Vitória/Derrota para os Pais Segundo Atletas do Alto Rendimento

    OpenAIRE

    Amblard, Isabela; Cruz, Fatima Leite

    2015-01-01

    Este estudo compreendeu as representações sociais da vitória/derrota para os pais segundo atletas-adolescentes do esporte de alto rendimento, na cidade do Recife. A adolescência é compreendida a partir da Psicologia Social-histórica, e o embasamento teórico-metodológico da Teoria das Representações Sociais abordou os sujeitos em diferentes contextos socioculturais, lugares de pertencimento, experiências, crenças, saberes e sentimentos compartilhados. Adotou-se a perspectiva pluri metodológica...

  15. Bird diversity and conservation of Alto Balsas (Southwestern Puebla, Mexico

    Directory of Open Access Journals (Sweden)

    Jorge E Ramírez-Albores

    2007-03-01

    Full Text Available Knowledge of the composition of the bird community in Alto Balsas (southwestern Puebla, Central Mexico is needed for management programs aiming at protection and conservation of bird species and their habitats I studied sites with tropical deciduous forest. Data were obtained during 1666 hours of field work in 238 days from March 1998 to September 2000. Six permanent transect (3.5 km long and 100 m wide; 30 to 40 ha in each transect were used to determine species richness in the study sites. The Shannon-Wiener diversity index was calculated for each site and Sorensen’s index was used to assess similarity between sites. One-way analysis of variance was used to test for differences between sites in species richness and diversity values. A total of 128 species were recorded, Tepexco (n = 75, H´= 3.76 and Puente Márquez (n = 61, H´= 3.62 were the sites that showed the greatest specific richness and diversity. However, species richness and diversity seasonally patterns were similar among sites (ANOVA p > 0.05, with highest diversity during the rainy season. Most species were resident; 42 were migrants. The avifauna was represented by 30 species associated with tropical deciduous forest and 12 from open habitats or heavily altered habitats. Insectivores were the best represented trophic category, followed by carnivores and omnivores. Rev. Biol. Trop. 55 (1: 287-300. Epub 2007 March. 31.Este estudio describe la diversidad avifaunística en sitios del Alto Balsas (suroeste de Puebla en el Centro de México y examina la variación en la diversidad de las especies de aves. El estudio fue llevado a cabo en sitios con presencia de bosque tropical caducifolio. Los datos fueron obtenidos durante 1666 horas de trabajo de campo en 238 días de Marzo 1998 a Septiembre 2000. Se realizaron seis transectos permanentes (de 3.5 km de longitud y 100 m de ancho; de 30 a 40 ha en cada transecto para determinar la riqueza de especies en los sitios de estudio. Se

  16. Mejoras tecnológicas en el proceso de inyección de carbón pulverizado en el horno alto

    Directory of Open Access Journals (Sweden)

    Babich, A.

    1996-04-01

    Full Text Available Blast furnace operation technology with pulverized coal injection (PCI has been carried out in order to replace the maximum amount of metallurgical coke in the burden. PC burning has been studied and methods and designs for the intensification of its combustion have been developed within the raceway. Recommendations for optimizing coal grinding have been developed. Problems of PC distribution inside the circumference of the furnace were investigated. Compensating technology for changes in reduction, heat exchange and other processes under PCI have been developed.

    Se estudia la tecnología de operación del horno alto con inyección de carbón pulverizado (ICP con el fin de sustituir el máximo posible de coque siderúrgico en la carga y se analizan aspectos de la combustión del carbón pulverizado (CP y los métodos y dispositivos elaborados para intensificar su combustión en la zona de toberas del horno alto. Se ofrecen recomendaciones para optimizar la molienda del carbón, se estudia la problemática de la distribución del CP a lo largo de la periferia del horno y se describe una tecnología elaborada para compensar las perturbaciones producidas por la inyección de CP en el horno alto sobre diferentes parámetros, tales como grado de reducción directa, intercambio térmico y otras.

  17. 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

  18. Environmental assessment for decontamination of the Three Mile Island Unit 2 reactor building atmosphere. Addendum 2. Draft NRC staff report for public comment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-04-01

    The reactor building purge system is an existing system originally installed for purging the reactor building atmosphere during normal operation or maintenance conditions. Use of the reactor building purge system in conjunction with the hydrogen control subsystem evaluated in Section 6.1 represents a variation in the purging alternative for decontaminating the Unit 2 reactor building atmosphere. This variation in the purging alternative would function only under meteorological conditions favorable to atmospheric dispersion. The reactor building purge system is capable of purging the building at flow rates of 5,000-50,000 cfm. Actual purge rates authorized during any time interval would be dependent on meteorological conditions and reactor building concentrations. Like the hydrogen control subsystem, this system would remove reactor building atmosphere through a filter system and discharge it through the 160-ft plant vent stack to the environment. The advantage of using the reactor building purge system in conjunction with the hydrogen control system is that it could decontaminate the reactor building atmosphere in a total elapsed purge time as short as approximately 5 days, as compared with the 60 days that would be required if the hydrogen purge subsystem were used alone. Use of this variation in the purge alternative would result in the release of radioactive materials to the environment. However, calculations based on actual meteorological and release-rate data would be used to monitor radioactive releases so that they do not exceed the requirements of 10 CFR Part 20, the design objectives of 10 CFR Part 50, Appendix 1 and the applicable requirements of 40 CFR 190.10.

  19. Re criticality assessment following reactor core damage in Fukushima unit 2

    International Nuclear Information System (INIS)

    Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

    2012-01-01

    Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

  20. RB reactor as the U-D2O benchmark criticality system

    International Nuclear Information System (INIS)

    Pesic, M.

    1998-01-01

    From a rich and valuable database fro 580 different reactor cores formed up to now in the RB nuclear reactor, a selected and well recorded set is carefully chosen and preliminarily proposed as a new uranium-heavy water benchmark criticality system for validation od reactor design computer codes and data libraries. The first results of validation of the MCNP code and adjoining neutron cross section libraries are resented in this paper. (author)