Protective effect of yerba mate intake on the cardiovascular system: a post hoc analysis study in postmenopausal women.

Science.gov (United States)

da Veiga, D T A; Bringhenti, R; Copes, R; Tatsch, E; Moresco, R N; Comim, F V; Premaor, M O

2018-01-01

The prevalence of cardiovascular and metabolic diseases is increased in postmenopausal women, which contributes to the burden of illnesses in this period of life. Yerba mate (Ilex paraguariensis) is a native bush from Southern South America. Its leaves are rich in phenolic components, which may have antioxidant, vasodilating, hypocholesterolemic, and hypoglycemic proprieties. This post hoc analysis of the case-control study nested in the Obesity and Bone Fracture Cohort evaluated the consumption of yerba mate and the prevalence of hypertension, dyslipidemia, and coronary diseases in postmenopausal women. Ninety-five postmenopausal women were included in this analysis. A questionnaire was applied to evaluate the risk factors and diagnosis of cardiovascular diseases and consumption of yerba mate infusion. Student's t-test and chi-square test were used to assess significant differences between groups. The group that consumed more than 1 L/day of mate infusion had significantly fewer diagnoses of coronary disease, dyslipidemia, and hypertension (P<0.049, P<0.048, and P<0.016, respectively). Furthermore, the serum levels of glucose were lower in the group with a higher consumption of yerba mate infusion (P<0.013). The serum levels of total cholesterol, LDL-cholesterol, HDL-cholesterol, and triglycerides were similar between the groups. This pragmatic study points out the benefits of yerba mate consumption for the cardiovascular and metabolic systems. The ingestion of more than 1 L/day of mate infusion was associated with fewer self-reported cardiovascular diseases and lower serum levels of glucose. Longitudinal studies are needed to evaluate the association between yerba mate infusion and reduction of cardiovascular diseases in postmenopausal women.

Effect of Chocolate and Yerba Mate Phenolic Compounds on Inflammatory and Oxidative Biomarkers in HIV/AIDS Individuals

OpenAIRE

Petrilli, Aline A.; Souza, Suelen J.; Teixeira, Andrea M.; Pontilho, Patricia M.; Souza, Jos? M. P.; Luzia, Liania A.; Rond?, Patricia H. C.

2016-01-01

Flavonoids in cocoa and yerba mate have a beneficial role on inflammation and oxidative disorders. Their effect on HIV individuals has not been studied yet, despite the high cardiovascular risk of this population. This study investigated the role of cocoa and yerba mate consumption on oxidative and inflammatory biomarkers in HIV+ individuals. A cross-over, placebo-controlled, double-blind, randomized clinical trial was conducted in 92 individuals on antiretroviral therapy for at least six mon...

Long-Term Dietary Supplementation with Yerba Mate Ameliorates Diet-Induced Obesity and Metabolic Disorders in Mice by Regulating Energy Expenditure and Lipid Metabolism.

Science.gov (United States)

Choi, Myung-Sook; Park, Hyo Jin; Kim, Sang Ryong; Kim, Do Yeon; Jung, Un Ju

2017-12-01

This study evaluated whether long-term supplementation with dietary yerba mate has beneficial effects on adiposity and its related metabolic dysfunctions in diet-induced obese mice. C57BL/6J mice were randomly divided into two groups and fed their respective experimental diets for 16 weeks as follows: (1) control group fed with high-fat diet (HFD) and (2) mate group fed with HFD plus yerba mate. Dietary yerba mate increased energy expenditure and thermogenic gene mRNA expression in white adipose tissue (WAT) and decreased fatty acid synthase (FAS) mRNA expression in WAT, which may be linked to observed decreases in body weight, WAT weight, epididymal adipocyte size, and plasma leptin level. Yerba mate also decreased levels of plasma lipids (free fatty acids, triglycerides, and total cholesterol) and liver aminotransferase enzymes, as well as the accumulation of hepatic lipid droplets and lipid content by inhibiting the activities of hepatic lipogenic enzymes, such as FAS and phosphatidate phosphohydrolase, and increasing fecal lipid excretion. Moreover, yerba mate decreased the levels of plasma insulin as well as the homeostasis model assessment of insulin resistance, and improved glucose tolerance. Circulating levels of gastric inhibitory polypeptide and resistin were also decreased in the mate group. These findings suggest that long-term supplementation of dietary yerba mate may be beneficial for improving diet-induced adiposity, insulin resistance, dyslipidemia, and hepatic steatosis.

Effect of Chocolate and Yerba Mate Phenolic Compounds on Inflammatory and Oxidative Biomarkers in HIV/AIDS Individuals.

Science.gov (United States)

Petrilli, Aline A; Souza, Suelen J; Teixeira, Andrea M; Pontilho, Patricia M; Souza, José M P; Luzia, Liania A; Rondó, Patricia H C

2016-05-23

Flavonoids in cocoa and yerba mate have a beneficial role on inflammation and oxidative disorders. Their effect on HIV individuals has not been studied yet, despite the high cardiovascular risk of this population. This study investigated the role of cocoa and yerba mate consumption on oxidative and inflammatory biomarkers in HIV+ individuals. A cross-over, placebo-controlled, double-blind, randomized clinical trial was conducted in 92 individuals on antiretroviral therapy for at least six months and at viral suppression. Participants were randomized to receive either 65 g of chocolate or chocolate-placebo or 3 g of yerba mate or mate-placebo for 15 days each, alternating by a washout period of 15 days. At baseline, and at the end of each intervention regimen, data regarding anthropometry, inflammatory, oxidative and immunological parameters were collected. High-sensitivity C-reactive protein, fibrinogen, lipid profile, white blood cell profile and thiobarbituric acid reactive substances were assessed. There was a difference between mean concentrations of HDL-c (ANOVA; p ≤ 0.05) among the different regimens: dark chocolate, chocolate-placebo, yerba mate and mate-placebo. When a paired Student t-test was used for comparisons between mean HDL-c at baseline and after each regimen, the mean concentration of HDL-c was higher after supplementation with dark chocolate (p = 0.008).

Effect of Chocolate and Yerba Mate Phenolic Compounds on Inflammatory and Oxidative Biomarkers in HIV/AIDS Individuals

Directory of Open Access Journals (Sweden)

Aline A. Petrilli

2016-05-01

Full Text Available Flavonoids in cocoa and yerba mate have a beneficial role on inflammation and oxidative disorders. Their effect on HIV individuals has not been studied yet, despite the high cardiovascular risk of this population. This study investigated the role of cocoa and yerba mate consumption on oxidative and inflammatory biomarkers in HIV+ individuals. A cross-over, placebo-controlled, double-blind, randomized clinical trial was conducted in 92 individuals on antiretroviral therapy for at least six months and at viral suppression. Participants were randomized to receive either 65 g of chocolate or chocolate-placebo or 3 g of yerba mate or mate-placebo for 15 days each, alternating by a washout period of 15 days. At baseline, and at the end of each intervention regimen, data regarding anthropometry, inflammatory, oxidative and immunological parameters were collected. High-sensitivity C-reactive protein, fibrinogen, lipid profile, white blood cell profile and thiobarbituric acid reactive substances were assessed. There was a difference between mean concentrations of HDL-c (ANOVA; p ≤ 0.05 among the different regimens: dark chocolate, chocolate-placebo, yerba mate and mate-placebo. When a paired Student t-test was used for comparisons between mean HDL-c at baseline and after each regimen, the mean concentration of HDL-c was higher after supplementation with dark chocolate (p = 0.008.

Anticonvulsant, neuroprotective and behavioral effects of organic and conventional yerba mate (Ilex paraguariensis St. Hil.) on pentylenetetrazol-induced seizures in Wistar rats.

Science.gov (United States)

Branco, Cátia Dos Santos; Scola, Gustavo; Rodrigues, Adriana Dalpicolli; Cesio, Verónica; Laprovitera, Mariajosé; Heinzen, Horacio; Dos Santos, Maitê Telles; Fank, Bruna; de Freitas, Suzana Cesa Vieira; Coitinho, Adriana Simon; Salvador, Mirian

2013-03-01

Epilepsy, which is one of the most common neurological disorders, involves the occurrence of spontaneous and recurrent seizures that alter the performance of the brain and affect several sensory and behavioral functions. Oxidative damage has been associated with post-seizure neuronal injury, thereby increasing an individual's susceptibility to the occurrence of neurodegenerative disorders. The present study investigated the possible anticonvulsive and neuroprotective effects of organic and conventional yerba mate (Ilex paraguariensis), a plant rich in polyphenols, on pentylenetetrazol (PTZ)-induced seizures in Wistar rats. The behavioral and polyphenolic profiles of the yerba mate samples were also evaluated. Infusions of yerba mate (50mg/kg) or distilled water were given to rats for fifteen days by oral gavage. On the 15th day the animals were subjected to open field test, and exploratory behavior was assessed. Subsequently, 60mg/kg PTZ (i.p.) was administered, and animals were observed for the appearance of convulsions for 30min. Latency for the first seizure, tonic-clonic and generalized seizures time, frequency of seizures and mortality induced by PTZ were recorded. The animals were then sacrificed, and the cerebellum, cerebral cortex and hippocampus were quickly removed and frozen to study the neuroprotective effects of yerba mate. The oxidative damage in lipids and proteins, nitric oxide levels, the activities of the antioxidant enzymes superoxide dismutase (Sod) and catalase (Cat) and non-enzymatic cellular defense (sulfhydryl protein) were quantified in all the tissues. The results showed that organic and conventional yerba mate infusions were able to reduce the frequency of seizures when compared to the PTZ group. Besides, organic yerba mate infusion decreases the tonic-clonic seizures time in relation to the PTZ group. It was also shown that organic and conventional yerba mate infusions reduced the oxidative damage in lipids and proteins and nitric oxide

CANDU fuel behaviour under LOCA conditions

International Nuclear Information System (INIS)

Kohn, E.

1989-07-01

This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

Dataset of the first transcriptome assembly of the tree crop “yerba mate” (Ilex paraguariensis and systematic characterization of protein coding genes

Directory of Open Access Journals (Sweden)

Patricia M. Aguilera

2018-04-01

Full Text Available This contribution contains data associated to the research article entitled “Exploring the genes of yerba mate (Ilex paraguariensis A. St.-Hil. by NGS and de novo transcriptome assembly” (Debat et al., 2014 [1]. By means of a bioinformatic approach involving extensive NGS data analyses, we provide a resource encompassing the full transcriptome assembly of yerba mate, the first available reference for the Ilex L. genus. This dataset (Supplementary files 1 and 2 consolidates the transcriptome-wide assembled sequences of I. paraguariensis with further comprehensive annotation of the protein coding genes of yerba mate via the integration of Arabidopsis thaliana databases. The generated data is pivotal for the characterization of agronomical relevant genes in the tree crop yerba mate -a non-model species- and related taxa in Ilex. The raw sequencing data dissected here is available at DDBJ/ENA/GenBank (NCBI Resource Coordinators, 2016 [2] Sequence Read Archive (SRA under the accession SRP043293 and the assembled sequences have been deposited at the Transcriptome Shotgun Assembly Sequence Database (TSA under the accession GFHV00000000.

Relacionando cultura y naturaleza

Directory of Open Access Journals (Sweden)

Thomas HEYD

2009-05-01

Full Text Available RESUMEN: En las diferentes culturas del mundo, el ambiente natural se percibe de diversas maneras, y en muchas sociedades no se considera como opuestos lo natural y lo cultural. En cuanto que la integridad del medio ambiente natural se ha convertido en algo muy preocupante, hay que preguntarse cómo concebir lo cultural en relación a lo natural para llegar a relacionarnos adecuadamente con la naturaleza. En este ensayo propongo que la naturaleza constituye una categoría importante y distintiva, que puede haber una cultura adecuada a la protección de, y al respeto por, la naturaleza, y que la administración y restauración de áreas naturales no son necesariamente contrarias al objetivo de preservar la naturaleza.ABSTRACT: The natural environment is perceived in distinctive ways in diverse cultures, and in many societies nature and culture are not considered as opposites. As the integrity of the natural environment becomes recognised as a major concern, we need to ask how culture and nature are, and can be, related, in order to determine what should be an appropiate relationship to nature. In this paper I propose that nature can be understood as an important, distinctive category, that a culture that respects and protects the natural environment is possible, and that the management and restoration of natural areas are not necessarily contradictory to the goal of preserving nature.

Physicochemical characterization of different trademarks of compound Yerba Maté and their herbs

Directory of Open Access Journals (Sweden)

Griselda Patricia Scipioni

2007-07-01

Full Text Available The objectives of this study were to evaluate the physicochemical characteristics of the main herbs used in the mixture of yerba maté with other aromatic herbs and the characterization of the trademarks of compound yerba maté. Moisture, water extract, total ash, acid-insoluble ash and caffeine concentration were determined. Results showed higher values of moisture content, total and aci-insoluble ash and lower water extracts in the herbs. Determinations were carried out in nine trademarks of compound yerba maté. In most cases they complied with the standards of the country with the exception of one trademark from Argentina.A erva-mate composta é um produto que se consome amplamente na região do MERCOSUL. Obtém-se misturando erva-mate com outras ervas aromáticas. O objetivo desta pesquisa foi o estudo das características físico-químicas das principais ervas usadas na mistura e a caracterização das marcas de erva-mate composta. Determinou-se a umidade, extração de água, cinzas totais, cinzas insolúveis ácidas e a concentração de cafeína. Encontraram-se nas ervas, valores padrões diferentes aos da erva-mate tais como valores maiores de conteúdo de umidade, cinzas totais, cinzas insolúveis ácidas e menores extratos de água. Fizeram-se determinações em nove marcas de erva-mate composta. Na maioria dos casos, cumpriam com as normas do país, exceto uma marca da Argentina

Study for Relation of Pressure and Aging Degradation during LOCA Test

International Nuclear Information System (INIS)

Kim, Jong Seog

2013-01-01

As result of this test, it was found that low pressure effect in aging was not significant compared with that of temperature. If temperature profile in LOCA test can satisfy the plant LOCA profile, no further analysis of pressure profile for aging degradation is necessary. For environmental qualification of electric equipment in containment building of nuclear power plant, LOCA test should be applied. During the LOCA test, temperature and pressure of LOCA chamber shall be controlled to meet a requirement of plant specific LOCA profile. It is general to keep LOCA test temperature and pressure above the plant specific LOCA profile. If the test temperature is lower than required profile in some time zone while it is higher in other time zone, calculation of total cumulated test temperature is required to compare with that of plant profile. Arrhenius equation can be applied for calculation of total temperature accumulation. If there is a deviation of pressure between test profile and plant specific profile, can we still use the same rule of temperature? Since the Arrhenius equation can't be applied to pressure, analysis of pressure effect to aging degradation is not easy. Study for relation of pressure and aging degradation during LOCA condition is described herein. To Study an aging degradation effect of pressure during LOCA test, comparison of IR during high LOCA pressure and low LOCA pressure were implemented. We expected low IR in high pressure because it contained a high concentration of oxygen which induces high aging degradation. Contrary to our expectation, IR of low pressure was lower than that of high pressure. It is assumed that high vibration of temperature profile to maintain the low pressure at high temperature induced supply of high enthalpy steam into LOCA chamber

Arsenic, cadmium and lead concentrations in Yerba mate commercialized in Southern Brazil by inductively coupled plasma mass spectrometry

Directory of Open Access Journals (Sweden)

Lisia Maria Gobbo dos Santos

2017-12-01

Full Text Available ABSTRACT: “Mate” or “Yerba Mate” (Ilex paraguariensis is a native South American plant, commonly consumed in Argentina, Paraguay, Uruguay and southern Brazil. Recent research has detected the presence of many vitamins and metals in this plant. Theses metals are also part of yerba mate’s mineral composition, due to soil and water contamination by pesticides and fertilizers, coal and oil combustion, vehicle emissions, mining, smelting, refining and the incineration of urban and industrial waste. Regardless of their origin, some inorganic elements, such as arsenic, cadmium and lead, are considered toxic, since they accumulate in all plant tissues and are, thus, introduced into the food chain. In this context, the aim of the present study was to determine and compare arsenic, cadmium, lead concentrations in 104 samples of yerba mate (Ilex paraguariensis marketed, and consumed in three southern Brazilian States, namely Paraná (PR, Santa Catarina (SC and Rio Grande do Sul (RS. Each element was determined by inductively coupled plasma mass spectrometry (ICP-MS, on a Nexion 300D equipment (Perkin Elmer. As, Cd and Pb concentrations in yerba mate leaves ranged from 0.015 to 0.15mg kg-1, 0.18 to 1.25mg kg-1 and 0.1 to 1.20mg kg-1, respectively. Regarding Cd, 84% of the samples from RS, 63% from PR and 75% from SC showed higher concentrations than the maximum permissible limit of 0.4mg kg-1 established by the Brazilian National Sanitary Surveillance Agency (ANVISA, while 7% of the samples from RS and 5% from PR were unsatisfactory for Pb. Concentrations were below the established ANVISA limit of 0.6mg kg-1 for all samples.

Design basis neutronics calculations for NRU-LOCA experiments

International Nuclear Information System (INIS)

Heaberlin, S.W.; Jenquin, U.P.; McNair, G.W.; Perry, R.T.; Trapp, T.J.; Zimmerman, M.G.

1979-08-01

The report describes the neutronics analysis for the LOCA simulation experiments in the NRU reactor. The experimental program will provide greater understanding of nuclear fuel assembly behavior during the heatup, reflood and quench sequence of a hypothetical LOCA. The decay heat and stored heat, which are the energy source in a LOCA will be simulated by fission heat provided by the NRU reactor. The reactor, the test and test operation are described

Safety studies on LOCA for N.S. Mutsu

International Nuclear Information System (INIS)

Kawasaki, Masayuki; Yaguchi, Shinnosuke

1978-01-01

A number of safety studies are under way concerning the reactor plant of N.S. Mutsu. One such study relates to Loss of Coolant Accidents (LOCA), which has been conducted to cover mainly the two subjects of experiments to ascertain the integrity of stainless steel fuel cladding under the action of the Emergency Core Cooling System (ECCS), and analysis of containment integrity following a LOCA. The stainless steel cladding tests were conducted to test swelling, rupture, oxidation and compression characteristics. Few reports are known to have been published in this domain, so that the present results should prove useful for future studies related to ECCS evaluation analyses on stainless steel fuel cladding. The containment integrity analysis covered variations of containment pressure and temperature following a LOCA, performed separately for short- and long-term periods. Estimates were also made on the changes in the hydrogen concentration present inside the containment after a LOCA. The results obtained should serve in determining the characteristic response to LOCA of marine reactor plants

Contributions to the quality control of two crops of economic importance : hops and yerba mate

NARCIS (Netherlands)

Wilson, Erica Georgina

2012-01-01

Quality control of plants is essential and at the same time very challenging.In this thesis, studies involving quality issues of two plants used in the production of two popular beverages, hops (in beer) and Ilex paraguariensis (yerba mate) were undertaken. Hops are used as bittering agents and to

Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

Energy Technology Data Exchange (ETDEWEB)

Nyman, R.; Erixon, S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B. [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B. [RSA Technologies, Visat, CA (United States)

1996-09-01

The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs.

Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

International Nuclear Information System (INIS)

Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

1996-09-01

The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs

Post-LOCA long term cooling performance in Korean standard nuclear power plants

International Nuclear Information System (INIS)

Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

1999-01-01

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break LOCA and large break LOCA. The RELAP5/MOD3.2.2 code is used to calculate the LTC sequences based on the LTC plan of the KSNPP. A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important manual action including the safety injection tank isolation in LTC procedure is investigated

LOCA and RIA studies at JAERI

International Nuclear Information System (INIS)

Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi

2004-01-01

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI. (Author)

LOCA analysis of the IRIS reactor

International Nuclear Information System (INIS)

Bajs, T.; Grgic, D.; Cavlina, N.

2003-01-01

The IRIS reactor (International Reactor Innovative and Secure) is an integral, light water cooled, medium power reactor. IRIS has been selected as an International Near Term Deployable (INTD) reactor, within the Generation IV International Forum activities. The IRIS concept addresses the key-requirements defined by the US DOE for next generation reactors, i.e. enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). An innovative safety approach has been developed to mitigate the IRIS response to small-to-medium Loss of Coolant Accident (LOCA). This strategy is based on the interaction of IRIS compact containment with the reactor vessel to limit initial blowdown, and on depressurization through the use of a passive Emergency Heat Removal System (EHRS). A small Automatic Depressurization System (ADS) provides supplementary depressurization capability. A pressure suppression system is provided to limit the pressure peak following the initial blowdown to well below the containment design limit. The ultimate result is that during a small-to-medium LOCA, the core remains covered for an extended period of time, without credit for emergency water injection or external core makeup. The IRIS LOCA response is based on 'maintaining water inventory' rather than on the principle of safety injection. This novel safety approach poses significant issues for computational and analysis methods since the IRIS vessel and containment are strongly coupled, and the system response is based on the interaction between the two. The small break LOCA was calculated using RELAP5/mod3.3 and GOTHIC codes. Break of the largest line connected to the IRIS Reactor Pressure Vessel (RPV) was analyzed. The results of the calculations confirmed good performance of the IRIS system during LOCA. (author)

Elite y naturaleza. ¿Naturaleza de elite?

Directory of Open Access Journals (Sweden)

Myriam Luz Jaramillo Giraldo

2005-04-01

Full Text Available El artículo da cuenta de las actitudes y representaciones de una colectividad particular con respecto a la naturaleza, su origen y consecuencias sobre el paisaje, la tierra, la estructura agraria, en la primera mitad del siglo XX –partiendo del XIX–. Ideas y representaciones de la elite que se encarga de “dirigir” el destino del país y que repercuten en todos los estratos, creando no sólo políticas y técnicas; también actitudes profundas y sus efectos visibles en los paisajes reales y en las mediaciones complejas de las representaciones fantásticas que los conforman.

MEDIOAMBIENTE, NATURALEZA Y ECOLOGÍA UN PROBLEMA RELACIONAL

Directory of Open Access Journals (Sweden)

Eliecer Mayorca Capataz

2016-08-01

Full Text Available Este trabajo es una reflexión para pensar la relación que el hombre en su caminar por la tierra ha mantenido con la naturaleza, la ecología y consigo mismo; relación que, en la escena global contemporánea, plantea desafíos y urgencias debido al abuso del dominio de poder del hombre, causando su destrucción sobre ella. Ante la evidencia y la envergadura de sus impactos nocivos la humanidad se ha empezado a cuestionar que no ha sido legítimo lo que se hace con la naturaleza, apareciendo la inquietud ética que moviliza hacia la necesidad nuevas formas de relación y cuidado. Para ello, se establece como objetivo realizar un análisis conceptual de la relación medio ambiente, naturaleza y ecología, en donde el hombre como actor, facilitador y responsable se estimule a adoptar una mejor comprensión, y construir un sistema de relaciones ecológico orientado a que los seres humanos se reconozcan parte integrante de la naturaleza, para superar la precariedad de la vida y ayude a colocarla en una posición privilegiada. Metodológicamente, el trabajo es de carácter documental bibliográfico, soportadas en investigaciones previas de Ewald (relación de saber y poder; Guardini (el ocaso de la modernidad Boff (en lenguaje asociativo de la naturaleza, Morín (ecosistemas en donde se articula análisis comprensivos de la relación hombre-naturaleza, y como reflexión se establece que, es desde el planteamiento de las relaciones como se puede cuidar nuestro ambiente.

LC-MSn study of the chemical transformations of hydroxycinnamates during yerba maté (Ilex paraguariensis) tea brewing.

Science.gov (United States)

Matei, Marius Febi; Jaiswal, Rakesh; Patras, Maria A; Kuhnert, Nikolai

2016-12-01

Yerba maté is one of the most popular beverages in South American countries and its consumption is associated with a wide array of health effects. In this study, we used advanced HPLC-ESI-MS n and HPLC-ESI-HRMS methods for the identification and characterization of hydroxycinnamates and their derivatives formed during the brewing process of yerba maté. We report on the hydroxylation of the hydroxycinnamates cinnamoyl substituent by conjugate addition of water to form 3-hydroxy-dihydrocinnamic acid derivatives using a series of model compounds, including caffeoylglucoses, dicaffeoylquinic acids, methyl caffeoylquinate and mono caffeoylquinic acids. The regiochemistry of conjugate addition was determined by targeted tandem MS experiments performed on authentic standards. It was interesting to note that hydroxylation of hydroxycinnamates produced cis and acyl-migration isomers, which is in line with previously reported data. Copyright © 2016. Published by Elsevier Ltd.

Estimation of LOCA break size using cascaded Fuzzy neural networks

Energy Technology Data Exchange (ETDEWEB)

Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

2017-04-15

Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA), which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict the progression of LOCAs in nuclear power plants. Therefore, accurate information on the LOCA break position and size should be provided to efficiently manage the accident. In this paper, the LOCA break size is predicted using a cascaded fuzzy neural network (CFNN) model. The input data of the CFNN model are the time-integrated values of each measurement signal for an initial short-time interval after a reactor scram. The training of the CFNN model is accomplished by a hybrid method combined with a genetic algorithm and a least squares method. As a result, LOCA break size is estimated exactly by the proposed CFNN model.

Leakage rate from LOCA-aged inflatable airlock seals Pickering NGS 'B' personnel doors

International Nuclear Information System (INIS)

Fayle, G.W.; Cordingley, D.C.

1985-01-01

In order to demonstrate to the Atomic Energy Control Board that an air-lock inflatable seal will function after a LOCA exposure, an inflatable seal intended for personnel doors at the Pickering NGS 'B' was exposed to the thermal/moisture conditions of the LOCA requirement. While attending to determine the post-LOCA leakage rate it was found that additional leaks developed during each post-LOCA inflation/deflation cycle. The seal had been significantly and irreparably deteriorated by the LOCA exposure. The test has demonstrated that this type of LOCA exposed seal should not be expected to withstand either additional pressure above 207 kPa or additional inflation/deflation cycling. A higher inflation pressure and/or cycling will reduce the likelihood of a post-LOCA seal retaining an inflation pressure sufficient to prevent leakage across the seal

Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

International Nuclear Information System (INIS)

Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

2010-01-01

To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

Analysis of factors affecting the LOCA test quality

International Nuclear Information System (INIS)

Wang Lu

2008-01-01

Localization of nuclear safety-related equipment has become an important way of nuclear power development in China. To meet this demand, the competence should be promoted in the following two areas, one is to develop the capability of R and D and manufacturing of nuclear safety-related equipment, the other is to implement equipment qualification according to relevant codes and standards. As LOCA test is one of the most important parts in the qualification test of nuclear safety-related equipment, the main factors related with the quality of the LOCA test are analyzed in this paper, and this may be a reference to improve the skills in designing, constructing and operating LOCA test devices. (authors)

Naturaleza de los derechos humanos

Directory of Open Access Journals (Sweden)

Carlos López Dawson

2016-07-01

Full Text Available En la formación de una nueva Carta Fundamental los constituyentes requerirán conocer o consensuar sobre la naturaleza de los derechos humanos; luego, determinar cómo el Estado hará efectivo el respeto de tales derechos. Para ello es necesario recurrir a la historia y evolución de estos derechos, y el presente trabajo trata de contribuir a un debate productivo eficiente sobre la naturaleza de los derechos humanos, para que los ciudadanos puedan decidir en el entendido de que se trata de una decisión democrática reflexionada y fundada en el humanismo. El análisis se sitúa en el contexto técnico-ideológico del sistema republicano vigente que corresponde al estado actual del derecho internacional.

Replacement divider plate performance under LOCA loading

International Nuclear Information System (INIS)

Huynk, H.M.; MClellan, G.H.; Schneider, W.G.

1997-01-01

A primary divider plate in a nuclear steam generator is required to perform its partitioning function with a minimum of cross leakage, without degradation in operating performance and without loss of structural integrity resulting from normal and accident loading. The design of the replacement divider plate for normal operating conditions is discussed in some detail in reference 1 and 2. This paper describes the structural response of the replacement divider plate to the severe loading resulting from a burst primary pipe. The loads for which the divider plate structural performance must be evaluated are mild to severe differential pressure transients resulting from several postulated sizes and types of pipe break scenarios. In the unlikely event of a severe Loss of Coolant Accident (LOCA) the divider plate or parts thereof must not exit the steam generator nor completely block the outlet nozzle. For the milder LOCA loads, the integrity of the divider plate and seat bars must be maintained. Analysis for the milder LOCA loads was carried out employing a conservative approach which ignores the actual interaction between the structure and the primary fluid. For these load cases it was shown that the divider plate does not become disengaged from the seat bars. For the more severe pipe breaks, the thermal-hydraulic analysis was coupled iteratively with the structural analysis, thereby taking into account divider plate deformation, in order to obtain a better prediction of the behaviour of the divider plate. In this manner substantial reduction in divider plate response to the more severe LOCA loading was achieved. It has been shown that, for the case of a postulated large LOCA (100% reactor inlet header), the disengagement of the divider plate from the seat bars resulted in an opening smaller than 1% of the divider plate area. (author)

Espiritualidades arbitrarias y naturaleza desnaturalizada

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Pablo López López

2017-08-01

Full Text Available Términos como «espiritualidad», «naturaleza», «naturalismo» y «religión» se han venido usando con frecuencia de modo engañoso desde el siglo decimoctavo, a partir del sector antirreligioso del Iluminismo británico y francés. De hecho, esos términos a veces significan lo contrario de lo que aparentemente expresan. Así pues, necesitamos explorar el significado real de dichos términos, en especial el de «naturalismo». También tenemos que analizar una serie de contradicciones acerca de la comprensión de nociones como «persona» y «religión». Nos proponemos argumentar una nueva fraternidad universal y personal basada en la real naturaleza humana, la real espiritualidad y la real religiosidad.

A novel procedure to measure the antioxidant capacity of Yerba maté extracts

Directory of Open Access Journals (Sweden)

Vanessa Graciela Hartwig

2012-03-01

Full Text Available Yerba maté extracts have in vitro antioxidant capacity attributed to the presence of polyphenolic compounds, mainly chlorogenic acids and dicaffeoylquinic acid derivatives. DPPH is one of the most used assays to measure the antioxidant capacity of pure compounds and plant extracts. It is difficult to compare the results between studies because this assay is applied in too many different conditions by the different research groups. Thus, in order to assess the antioxidant capacity of yerba maté extracts, the following procedure is proposed: 100 µL of an aqueous dilution of the extracts is mixed in duplicate with 3.0 mL of a DPPH 'work solution in absolute methanol (100 µM.L-1, with an incubation time of 120 minutes in darkness at 37 ± 1 °C, and then absorbance is read at 517 nm against absolute methanol. The results should be expressed as ascorbic acid equivalents or Trolox equivalents in mass percentage (g% dm, dry matter in order to facilitate comparisons. The AOC of the ethanolic extracts ranged between 12.8 and 23.1 g TE % dm and from 9.1 to 16.4 g AAE % dm. The AOC determined by the DPPH assay proposed in the present study can be related to the total polyphenolic content determined by the Folin-Ciocalteu assay.

Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions

International Nuclear Information System (INIS)

Pettersson, Kjell; Chung, Haijung; ); Billone, Michael; Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Hache, George; Papin, Joelle; Heins, Lothar; Hozer, Zoltan; In de Betou, Jan; Kelppe, Seppo; Mayer, Ralph; Scott, Harold; Voglewede, John; Sonnenburg, Heinz; Sunder, Sham; Valach, Mojmir; Vrtilkova, Vera; Waeckel, Nicolas; Wiesenack, Wolfgang; Zimmermann, Martin

2009-01-01

The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the current understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burn-up and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including as regards experimental approaches and interpretation and the use of experimental data relevant for safety. In 1986, a working group of the NEA Committee on the Safety of Nuclear Installations (CSNI) issued a state-of-the-art report on water reactor fuel behaviour in design-basis accident (DBA) conditions. The 1986 report was limited to the oxidation, embrittlement and deformation of pressurised water reactor (PWR) fuel in a loss-of-coolant accident (LOCA). Since then, considerable experimental and analytical work has been performed, which has led to a broader and deeper understanding of LOCA-related phenomena. Further, new cladding alloys have been produced, which might behave differently than the previously used Zircaloy-4, both under normal operating conditions and during transients. Compared with 20 years ago, fuel burn-up has been significantly increased, which requires extending the LOCA database in order to cover the high burnup range. There was also a clear need to address LOCA performance for reactor types other than PWRs. The present report has been prepared by the WGFS and covers the following technical aspects: - Description of different LOCA scenarios for major types of reactors: BWRs, PWRs, VVERs and to a lesser extent CANDUs. - LOCA phenomena: ballooning, burst, oxidation, fuel relocation and possible fracture at quench. - Details of high-temperature oxidation behaviour of various cladding materials. - Metallurgical phase change, effect of hydrogen and oxygen on residual cladding ductility. - Methods for LOCA testing, for example two-sided oxidation and ring compression for ductility, and integral quench test for

Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

International Nuclear Information System (INIS)

Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

2017-01-01

Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

Energy Technology Data Exchange (ETDEWEB)

Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

2017-04-15

Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

NPP Krsko small break LOCA analysis

International Nuclear Information System (INIS)

Mavko, B.; Petelin, S.; Peterlin, G.

1987-01-01

Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

Synthesis of the works of geophysical (electrical soundings) carried out in the region of the anomalies of Yerba Sola (Cerro Largo) Uruguay

International Nuclear Information System (INIS)

Perrin, J.

1983-01-01

In this report has been studied geographical localization, geology, geophysics prospection, equipment and methods used and their results in Yerba Sola anomalies Province Cerro Largo and Fraile Muerto formation

Intensity of bitterness of processed yerba mate leaves originated in two contrasted light environments

Directory of Open Access Journals (Sweden)

Miroslava Rakocevic

2008-06-01

Full Text Available The bitterness intensity of beverage prepared from the leaves produced on the males and females of yerba mate (Ilex paraguariensis, grown in the forest understory and monoculture, was evaluated. The leaves were grouped by their position (in the crown and on the branch tips and by the leaf age. The leaf gas exchange, leaf temperature and photosynthetic photon flux density were observed. Inter and intra-specific competition for light and self-shading showed the same effect on yerba mate beverage taste. All the shading types resulted in bitterer taste of the processed yerba mate leaves compared to the leaves originated under the direct sun exposure. The leaves from the plants grown in the monoculture showed less bitterness than those grown in the forest understory. This conclusion was completely opposite to the conventionally accepted paradigm of the yerba mate industries. The leaves from the tips (younger leaves of the plants grown in the monoculture resulted a beverage of softer taste; the males produced less bitter leaves in any light environment (forest understory or in the crown in monoculture. The taste was related to the photosynthetic and transpiration rate, and leaf temperature. Stronger bitterness of the leaves provided from the shade conditions was related to the decreased leaf temperature and transpiration in the diurnal scale.Mediu-se a intensidade de amargor da bebida preparada a partir de folhas da erva-mate (Ilex paraguariensis de diversas idades, situadas em duas posições na copa (interior e ponteiras, produzidas por plantas masculinas e femininas cultivadas na floresta antropizada e em monocultura. As trocas gasosas foliares, a temperatura de folhas e a densidade de fluxo de fótons fotossinteticamente ativos também foram medidas. Com isso verificou-se que a idéia corrente de que o sombreamento está diretamente relacionado ao sabor suave do chimarrão é completamente equivocada, já que as competições inter- e intra

Prediction of LOCA Break Size Using CFNN

Energy Technology Data Exchange (ETDEWEB)

Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Kim, Dong Yeong; Na, Man Gyun [Chosun University Gwangju (Korea, Republic of)

2016-05-15

The NPPs have the emergency core cooling system (ECCS) such as a safety injection system. The ECCS may not function properly in case of the small break size due to a slight change of pressure in the pipe. If the coolant is not supplied by ECCS, the reactor core will melt. Therefore, the meltdown of reactor core have to be prevented by appropriate accident management through the prediction of LOCA break size in advance. This study presents the prediction of LOCA break size using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model is a data-based method that requires data for its development and verification. The data were obtained by numerically simulating severe accident scenarios of the optimized power reactor (OPR1000) using MAAP code, because real severe accident data cannot be obtained from actual NPP accidents. The CFNN model has been designed to rapidly predict the LOCA break size in LOCA situations. The CFNN model was trained by using the training data set and checked by using test data set. These data sets were obtained using MAAP code for OPR1000 reactor. The performance results of the CFNN model show that the RMS error decreases as the stage number of the CFNN model increases. In addition, the performance result of the CFNN model presents that the RMS error level is below 4%.

ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-06-01

Full Text Available ABSTRAK ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA. Kecelakaan yang diakibatkan oleh kehilangan pendingin (loss of coolant accident / LOCA dari sistem reaktor merupakan kejadian dasar desain yang tetap diantisipasi dalam desain reaktor daya yang mengadopsi teknologi Generasi II hingga IV. LOCA ukuran kecil (small break LOCA memiliki dampak yang lebih signifikan terhadap keselamatan dibandingkan LOCA ukuran besar (large break LOCA seperti terlihat pada kejadian Three-Mile Island (TMI. Fokus makalah adalah pada analisis small break LOCA pada reaktor daya maju Generasi III+ yaitu AP1000 dengan mensimulasikan tiga kejadian pemicu yaitu membukanya katup Automatic Depressurization System (ADS secara tak disengaja, putusnya salah satu pipa Direct Vessel Injection (DVI secara double-ended, dan putusnya pipa lengan dingin dengan diameter bocoran 10 inci. Metode yang digunakan adalah simulasi kejadian pada model AP1000 yang dikembangkan secara mandiri menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Dampak yang ingin dilihat adalah kondisi teras selama terjadinya small break LOCA yang terdiri dari pembentukan mixture level dan transien temperatur kelongsong bahan bakar. Hasil simulasi menunjukkan bahwa mixture level untuk semua kejadian small break LOCA berada di atas tinggi teras aktif yang menunjukkan tidak terjadinya core uncovery. Adanya mixture level berpengaruh pada transien temperatur kelongsong yang menurun dan menunjukkan pendinginan bahan bakar yang efektif. Hasil di atas juga identik dengan hasil perhitungan program lain yaitu NOTRUMP. Keefektifan pendinginan teras juga disebabkan oleh berfungsinya injeksi pendingin melalui fitur keselamatan pasif yang menjadi ciri reaktor daya AP1000. Secara keseluruhan, hasil analisis menunjukkan model AP1000 yang telah dikembangkan dengan RELAP5 dapat digunakan untuk keperluan analisis kecelakaan dasar desain pada reaktor daya maju AP1000. Kata kunci: analisis

Audit calculation for the LOCA methodology for KSNP

Energy Technology Data Exchange (ETDEWEB)

Lee, Un Chul; Park, Chang Hwan; Choi, Yong Won; Yoo, Jun Soo [Seoul National Univ., Seoul (Korea, Republic of)

2006-11-15

The objective of this research is to perform the audit regulatory calculation for the LOCA methodology for KSNP. For LBLOCA calculation, several uncertainty variables and new ranges of those are added to those of previous KINS-REM to improve the applicability of KINS-REM for KSNP LOCA. And those results are applied to LBLOCA audit calculation by statistical method. For SBLOCA calculation, after selecting BATHSY9.1.b, which is not used by KHNP, the results of RELAP5/Mod3.3 and RELAP5/MOD3.3ef-sEM for KSNP SBLOCA are compared to evaluate the conservativeness or applicability of RELAP5/MOD3.3ef-sEM code for KSNP SBLOCA. The result of this research can be used to support the activities of KINS for reviewing the LOCA methodology for KSNP proposed by KHNP.

Accomplishments of LOCA/ECCS experimental research at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Tasaka, Kanji; Murao, Yoshio; Koizumi, Yasuo

1984-01-01

Japan Atomic Energy Research Institute has investigated loss-of-coolant accident (LOCA)/emergency core cooling system (ECCS) from 1970. Major results of the LOCA/ECCS research are summarized in this report. ROSA-II program was LOCA/ECCS research for a pressurized water reactor (PWR) and ROSA-III program was for a boiling water reactor (BWR). The both test facilities were scaled at approximately 1/400 of the respective reference PWR and BWR. Large scale reflood test is research on reflood phenomena during a large break LOCA of PWR. The test facility is scaled at approximately 1/20 of the reference PWR and the research is still being continued. (author)

Taipower's approach in development of in-house LOCA analysis capability

International Nuclear Information System (INIS)

Wang, L.C.

2004-01-01

One lesson learned from the Three Mile Island (TMI) accident was the analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or nuclear fuel suppliers for small break Loss Of Coolant Accident (LOCA) analysis for compliance with appendix K to 10CFR50 should be revised, so, a technology transfer program and a training program of a new LOCA analysis methodology for Taipower's engineers is briefly described in this paper. Also, an other lesson learned from the TMI accident was the plant-specific calculations using NRC-approved models for small-break LOCA to show compliance with 10CFR50.46 should be submitted for NRC approval, so, a study by Taiwan Power Company (TPC) under the guidance of Yankee Atomic Electric Company (YAEC) has been undertaken to perform this analysis for Maanshan nuclear power plant. The results of the 4 inch line break LOCA analysis is described in this paper. (author)

An IPSN research programme to resolve pending LOCA issues

International Nuclear Information System (INIS)

Mailliat, A.; Grandjean, C.; Clement, B.

2001-01-01

Studies performed in IPSN and elsewhere pointed out that high burnup may induce specific effects under LOCA conditions, especially those related with fuel relocation. Uncertainties exist regarding how much these effects might affect the late evolution of the accident transient and the associated safety issues. IPSN estimates that a better knowledge of specific phenomena is required in order to resolve the pending uncertainties related to LOCA criteria. IPSN is preparing the so called APRP-Irradie (High Burnup fuel LOCA) programme. One of the important aspect of this programme is in-pile experiments involving bundle geometries in the PHEBUS facility located at Cadarache, France. A feasibility study for such an experimental programme is underway and should provide soon, a finalized project including cost and schedule aspects. (authors)

NI NATURALEZA NI AMBIENTE: EL OPERACIONALISMO SISTÉMICO Y EL ECOLOGISMO

Directory of Open Access Journals (Sweden)

GÓMEZ ECLIEVERRI LUIS FEMANDO

2010-05-01

Full Text Available Algunos críticos de la presente crisis de la modernidad ortodoxa que se ubican desde el ambientalismo o el ecologismo, ven en el concepto de naturaleza una de las raíces del actual estado de cosas. Por este motivo, creemos que el ecologismo como posición que comparte muchas de estas críticas, debería construirse a partir de prácticas discursivas que no tengan a la idea de naturaleza como uno de sus conceptos o creencias centrales. Con este objetivo, presentamos al operacionalismo como alternativa teórica para el ecologismo, pues no tiene como uno de sus fundamentos el concepto de naturaleza, ni lo sustituye con términos como ambiente, el cual parece mantener algunas de las dificultades que los críticos han visto en el concepto de naturaleza.

Bio-mechanical assessment toward throwing and lifting process of i-LOCA (Innovative Lobster Catcher)

Science.gov (United States)

Sudiarno, A.; Dewi, D. S.; Putri, M. A.

2018-04-01

Indonesia is the country rich in marine resource, one of which is lobster. East java, one of Indonesian province, especially in Region of Gresik and Lamogan, has very huge potential of lobster. Current condition shown that lobster catch by the fisherman mostly depend on lucky factor, which the lobster unintentionally trapped in fisherman’s fish net. By using this mechanism, the number of lobster catch cannot be optimum. Previous researches have produced two versions of i-LOCA, Innovative Lobster Catcher, a special tool for catching the lobster. Although produce more lobster catch, second version of i-LOCA still needs to be scrutinized, one of that is bio-mechanical assessment. The second version of i-LOCA still has no tool to ease throwing and lifting it into the sea. This condition cause Musculoskeletal Disorder (MSD) toward the fisherman. This research perform bio-mechanical assessment toward throwing and lifting process in order to suggest improvement for i-LOCA as the third version. Based on body moment calculation, we found that throwing and lifting process of third version of i-LOCA, each was 3 times and 2 times better than second version of i-LOCA. Meanwhile, Rapid Entire Body Assessment (REBA) score of throwing and lifting process for third version of i-LOCA can be reduced by 5 points compared to second version of i-LOCA.

Preliminary Results Of LOCA Problem For APR1400 Reactor

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Le Thi Thu; Vo Thi Huong; Le Van Hong

2011-01-01

Several features of NPP with APR1400 nuclear reactor during a loss of coolant accident (LOCA) are investigated in this study. The report describes some main design characteristics of an engineering safety systems of APR1400 and the thermal hydraulic calculation results for steady-state using MARS and RELAP/SCDAPSIM codes. Large Break LOCA accident has been analyzed and evaluated based on acceptable criteria for ECCS given by US NRC. The results from cold leg break LOCA with broken area of 0.0465 m 2 in case of high pressure safety injection system (HPSI) failed to operate or 2 and 4 HPSI pumps are activated. The preliminary results of this work is a part of collaboration between INST researchers and KAERI experts in using RELAP tool for safety analysis of NPPs. (author)

Sobre naturaleza e historia en el Humanismo español

Directory of Open Access Journals (Sweden)

Maravall, José Antonio

2003-04-01

Full Text Available Lo que el hombre, en cada momento de su existir, piensa que son las cosas físicas que hay en su mundo en torno constituye, sin duda, una de las cosas, a su vez, más importantes que posee. Ahora bien: eso que el hombre cree que son las cosas que le rodean, formulado con un sentido de unidad, significa su concepto de naturaleza. Por ende, su concepto de naturaleza representa algo decisivo para el hombre y muy interesante, por esa sola razón, para poder entender el tipo humano que en cada época se nos manifiesta. Podría hacerse una sumamente sugestiva historia de la cultura, siguiendo el hilo de los cambios en el concepto de naturaleza.

Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

International Nuclear Information System (INIS)

Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick; Rudland, Dave

2010-12-01

The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

Energy Technology Data Exchange (ETDEWEB)

Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick (Battelle Columbus (United States)); Rudland, Dave (Nuclear Regulatory Commission (United States))

2010-12-15

The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

Effect of oxygen in the simulated LOCA environments of the degradation of cable insulating materials

International Nuclear Information System (INIS)

Kusuma, Y.; Okada, S.; Itoh, M.; Yagi, T.; Yoshikawa, M.; Yoshida, K.; Machi, S.; Tamura, N.; Kawakami, W.

1990-01-01

Five kinds of insulating and jacketing materials for the cables used in nuclear power plants were exposed to various LOCA environments of both simultaneous and sequential methods using SEAMATE-II. Experimental conditions of the simultaneous LOCA tests were done at different radiation dose rate, steam temperature and amount of air added to the LOCA environments. The sequential tests consist of two stages, that is, pre-irradiation and subsequent steam/spray exposure. Pre-irradiation conditions and subsequent steam/spray exposure conditions of the sequential LOCA tests are systematically changed in order to find appropriate conditions which can bring about the degradation of same degree to those obtained for various simultaneous LOCA simulations. Tensile properties, insulating resistance and water sorption of the insulating materials exposed to various LOCA environments are measured and discussed. (author). 11 refs, 19 figs, 3 tabs

MELCOR ex-vessel LOCA simulations for ITER+

International Nuclear Information System (INIS)

Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

1995-01-01

Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

Status of efforts to evaluate LOCA frequency estimates using combined PRA and PFM approaches

International Nuclear Information System (INIS)

Wilkowski, G.; Rudland, D.; Tregoning, R.; Scott, P.

2002-01-01

The risk-informed reevaluation of 10 CFR 50.46 (along with Appendix K and GDC 35), the emergency core cooling system (ECCS) requirements, utilizes loss of coolant accident (LOCA) initiating event frequencies to evaluate the technical basis for potential related rule changes. A longer-term effort is considering redefining the maximum design basis pipe break size for sizing the ECCS system. In the past few years, the U.S. Nuclear Regulatory Commission (NRC) has utilized NUREG/CR-5750 pipe-break LOCA estimated for initiating event frequencies. However, several failure mechanisms have recently emerged at plants which have not been evident within the service period covered by the NUREG/CR-5750 estimates. The concern is that these and other potential aging-related mechanisms may not be adequately represented within the NUREG/CR-5750 LOCA estimates. Additionally, LOCAs can occur from failure of active components (e.g. safety relief valves, reactor coolant pump seals, etc.) and other non-pipe break passive failures (e.g. steam generator tubes). The LOCA contributions from these additional sources must also be considered in deciding the design basis break size. The LOCA estimates must also attempt to capture expected future changes in the LOCA frequencies so that the estimates are pertinent up through the end of the license renewal period. (orig.)

Evaluation of post-LOCA long term cooling performance in Korean standard nuclear power plants

International Nuclear Information System (INIS)

Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

2001-01-01

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT) isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences. (author)

Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

International Nuclear Information System (INIS)

Muhammad Subekti

2009-01-01

Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

La naturaleza como contraseña del comportamiento moral en Lucrecio

Directory of Open Access Journals (Sweden)

Ramón Román Alcalá

2015-01-01

Full Text Available Voy a intentar probar que en Lucrecio la naturaleza actúa como elemento clave para todo su sistema. Con una lógica coherente el poeta demuestra que la naturaleza como principio de realidad está sustentando, primero la creación del cosmos como principio material, segundo, el desarrollo de la raza humana, como principio antropológico y, por último, la acción y la conducta humana como principio moral. La naturaleza es, así, una marca para la comprensión del cosmos, una clave para el conocimiento de la humanidad y una contraseña que activa el comportamiento moral.

Progress in realistic LOCA analysis

Energy Technology Data Exchange (ETDEWEB)

Young, M Y; Bajorek, S M; Ohkawa, K [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

1994-12-31

While LOCA is a complex transient to simulate, the state of art in thermal hydraulics has advanced sufficiently to allow its realistic prediction and application of advanced methods to actual reactor design as demonstrated by methodology described in this paper 6 refs, 5 figs, 3 tabs

Benchmark Calculations on Halden IFA-650 LOCA Test Results

International Nuclear Information System (INIS)

Ek, Mirkka; Kekkonen, Laura; Kelppe, Seppo; Stengaard, J.O.; Josek, Radomir; Wiesenack, Wolfgang; Aounallah, Yacine; Wallin, Hannu; Grandjean, Claude; Herb, Joachim; Lerchl, Georg; Trambauer, Klaus; Sonnenburg, Heinz-Guenther; Nakajima, Tetsuo; Spykman, Gerold; Struzik, Christine

2010-01-01

The assessment of the consequences of a loss-of-coolant accident (LOCA) is to a large extent based on calculations carried out with codes especially developed for addressing the phenomena occurring during the transient. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have not only generated a need to re-examine the LOCA safety criteria and to verify their continued validity, but also to confirm that codes show an appropriate performance especially with respect to high burnup phenomena influencing LOCA fuel behaviour. As part of international efforts, the OECD Halden Reactor Project program implemented a test series to address particular LOCA issues. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The Halden LOCA series, using high burnup fuel segments, contains test cases well suited for checking the ability of LOCA analysis codes to predict or reproduce the measurements and to provide clues as to where the codes need to be improved. The NEA Working Group on Fuel Safety, WGFS, therefore decided to conduct a code benchmark based on the Halden LOCA test series. Emphasis was on the codes' ability to predict or reproduce the thermal and mechanical response of fuel and cladding. Before starting the benchmark, participants were given the opportunity to tune their codes to the experimental system applied in the Halden LOCA tests. To this end, the data from the two commissioning runs were made available. The first of these runs went

Nota científica: perfil bioquímico de ratos alimentados com iogurte contendo extrato de erva-mate (Ilex paraguariensis St. Hil Scientific Note: biochemical profile of rats fed yogurt containing yerba mate (Ilex paraguariensis St. Hil extract

Directory of Open Access Journals (Sweden)

Franciele Taís Ril

2011-12-01

Full Text Available Este estudo teve como objetivo avaliar o efeito do iogurte contendo extrato de erva-mate (Ilex paraguariensis e com/sem culturas probióticas sobre o perfil lipídico, glicêmico, hepático e renal de ratos alimentados com esses tipos de iogurtes. Ratos da linhagem Wistar (42 foram divididos em três grupos (n=14, e receberam iogurte sem extrato de erva-mate, iogurte com extrato de erva-mate 0,1% e iogurte com extrato de erva-mate 0,1% e culturas probióticas, durante 30 dias. Não foi observado no presente estudo efeito significativo do extrato de erva-mate sobre os níveis de colesterol total, colesterol HDL, triglicerídeos, uréia, ácido úrico, creatinina, glicose e na atividade das enzimas fosfatase alcalina, aspartato aminotransferase e alanina aminotransferase. O extrato de erva-mate, 0,1% no iogurte, não interfere no metabolismo de ratos alimentados por 30 dias.The objective of this study was to evaluate the effect of yoghurt containing yerba-mate extract (Ilex paraguariensis, with and without probiotic cultures, on the lipidic, glycemic, hepatic and kidney profiles of rats fed these types of yoghurt. Wistar rats (42 were divided into three groups (n=14 and for 30 days were fed yoghurt without yerba-mate extract, yoghurt with 0.1% yerba-mate extract and yoghurt with 0.1% yerba-mate extract and probiotic cultures. No significant effect of the yerba mate extract on the levels of total cholesterol, HDL cholesterol, triglycerides, urea, uric acid, creatinine, glucose and the activity of the enzymes alkaline phosphatase, aspartate aminotransferase and alanine aminotransferase was observed in the present study. The addition of 0.1% yerba mate extract to the yoghurt did not interfere with the metabolism of the rats during 30 days.

In-core LOCA (PTR) analysis with poisoned moderator

International Nuclear Information System (INIS)

Kim, S. R.; Kim, B. G.; Kim, T. M.; Choi, J. H.; Kim, Yun Ho; Choi, Hoon

2005-01-01

CANDU reactors have been analyzed and evaluated for the postulated in-core LOCA while the reactor is operating normally with low moderator poison concentration. However, when the reactor is operating with relatively large amounts of boron and/or gadolinium poisons in the moderator, the assessment for fuel integrity was required for pressure tube rupture (PTR) accident. The methodology of in-core LOCA analysis with poisoned moderator is developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for CANDU reactor recently. The developed methodology and results are presented

Analysis for Passive Safety Injection of IPSS in Various LOCAs

International Nuclear Information System (INIS)

Kim, Sangho; Chang, Soonheung

2013-01-01

The Fukushima accident shows US the possibility of accidents that are beyond a designed imagination. Lots of lessons can be shortly summarized into three issues. First of all, the original cause was the occurrence of a Station Black-Out (SBO). Even if engineers considered the possibility of a loss of offsite power enough to be managed, the failure of EDGs seemed to be unnoticed. The second is poor operation and accident management. They could not understand the overall system and did not check the availability of alternating systems. The third is the large release of radioactive materials outside the containment. Even if SBO occurred and the accident was not managed well, all the means must have prevented the large release out of containment. After that, lots of problems were pointed and numerous actions were carried out in each country. The representative proposals are AAC, additional physical barrier, bunker concept and large big tank. Integrated passive safety system (IPSS) was proposed as one of the solutions for enhancing the safety. IPSS can cope with a SBO and accidents with a SBO. IPSS has five functions which are passive decay heat removal, passive safety injection, passive containment cooling, passive in-vessel retention and filtered venting system. The results showed a high performance of removing decay heat through steam generator cooling by forming natural circulation in the primary circuit. The design concept of passive safety injection system (PSIS) consists of the injection line from integrated passive safety tank (IPST) to reactor vessel. The previous works were only focused on a double ended guillotine break LOCA in SBO. The purpose of this paper is to analyze the performance of PSIS in IPSS for various LOCAs by using MARS (Multi-dimensional Analysis of Reactor Safety) code. The simulated accidents were LOCAs which were accompanied with a SBO. The conditions of the LOCAs were varied only for the size of break. It shall show the capability of PSIS

LA FARMACIA NATURALEZA – FUENTE DE FÁRMACOS EN EL SIGLO XXI

OpenAIRE

Oliver Callies

2011-01-01

La naturaleza ha sido fuente de fármacos desde el inicio de la historia y en muchas culturas lo sigue siendo hoy en día, como en la medicina tradicional china o la ayurveda. Por otra parte, un 50% de los fármacos actualmente usados en la medicina del mundo occidental derivan directa o indirectamente de la naturaleza, como la Aspirina® o el Taxol®. El camino que una sustancia recorre desde la naturaleza hasta convertirse en un nuevo fármaco es largo, laborioso y, en general, muy costoso. Los p...

Review of high burn-up RIA and LOCA database and criteria

International Nuclear Information System (INIS)

Vitanza, C.; Hrehor, M.

2006-01-01

This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

International Nuclear Information System (INIS)

Liang, Thomas K.S.; Chou, Ling-Yao; Zhang, Zhongwei; Hsueh, Hsiang-Yu; Lee, Min

2011-01-01

Highlights: → A new LOCA licensing methodology (DRHM, deterministic-realistic hybrid methodology) was developed. → DRHM involves conservative Appendix K physical models and statistical treatment of plant status uncertainties. → DRHM can generate 50-100 K PCT margin as compared to a traditional Appendix K methodology. - Abstract: It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.

Yerba Mat? (Ilex paraguariensis) Metabolic, Satiety, and Mood State Effects at Rest and during Prolonged Exercise

OpenAIRE

Alkhatib, Ahmad; Atcheson, Roisin

2017-01-01

Yerba Mat? (YM), has become a popular herb ingested for enhancing metabolic health and weight-loss outcomes. No studies have tested the combined metabolic, satiety, and psychomotor effects of YM during exercise. We tested whether YM ingestion affects fatty acid oxidation (FAO), profile of mood state score (POMS), and subjective appetite scale (VAS), during prolonged moderate exercise. Twelve healthy active females were randomized to ingest either 2 g of YM or placebo (PLC) in a repeated-measu...

PERSPECTIVA DE NIÑOS Y NIÑAS DE TRANSICIÓN SOBRE NATURALEZA

Directory of Open Access Journals (Sweden)

Natalia Silva Garnica

2015-12-01

Full Text Available Se exploró la perspectiva de niños y niñas de transición sobre naturaleza.   Más específicamente se indagó cómo definen la naturaleza y el medio ambiente y qué sentimientos les genera. Participaron 49 niños y niñas entre los cinco y siete años de edad de dos instituciones de educación formal en el contexto urbano y se usaron diversas  herramientas  que  atendieron  la  expresión  oral,  escrita  y  gráfica.    Los resultados  muestran  que  los  niños  y  las  niñas  de  transición  tienden  a  definir  la naturaleza como elementos aislados, en su mayoría bióticos, que es confuso para ellos si los seres humanos hacemos parte de la naturaleza, que para ellos la naturaleza está a la vez cerca pero lejos y que medio ambiente es un concepto difícil de definir a esta edad. Manifestaron  sentimientos  positivos  frente  a  la  naturaleza  (emoción,  curiosidad, alegría, evidenciando biofilia en la primera infancia en el contexto citadino.   Las herramientas que capturan la voz de los niños resultaron siendo muy informativas pues muestran una perspectiva más dinámica y compleja.   Los resultados demuestran que niños y niñas de transición tienen una perspectiva clara sobre naturaleza y señalan algunos aspectos por fortalecer, en particular abordar más explícitamente la diversidad biológica y ambiental y reconocer que los humanos somos parte integral de ella y lo que esto implica

Por una estética de la naturaleza: la belleza natural como argumento ecologista

Directory of Open Access Journals (Sweden)

Tafalla, Marta

2005-06-01

Full Text Available In this paper, I defend the thesis that the aesthetics of nature can offer to the ethics of nature one of the best ecological arguments. I believe that !he aesthetic value of nature, which was traditionally neglected by philosophy, can revitalize the discussions which take place in the ethics of nature, and allows to approach in a new way the question if nature has an intrinsic or an instrumental value.

En este artículo defiendo la tesis de que la estética de la naturaleza puede ofrecer a la ética de la naturaleza uno de los mejores argumentos ecologistas. Creo que el valor estético de la naturaleza, tradicionalmente olvidado por la filosofía, puede revitalizar las discusiones que tienen lugar en la ética de la naturaleza, y permite enfocar de un modo nuevo la cuestión de si la naturaleza posee un valor intrínseco o un valor instrumental.

Cobalt-60 simulation of LOCA [loss of coolant accident] radiation effects

International Nuclear Information System (INIS)

Buckalew, W.H.

1989-07-01

The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs

Qinshan NPP large break LOCA safety analysis

International Nuclear Information System (INIS)

Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

1997-01-01

Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

TRANSFORMACIONES EN LAS INVESTIGACIONES ANTROPOLÓGICAS sobre naturaleza, ecología y medio ambiente

Directory of Open Access Journals (Sweden)

ASTRID ULLOA

2001-01-01

Full Text Available EN LAS INVESTIGACIONES ANTROPOLÓGICAS HA HABIDO UN LARGO PROCESO DE TRANSformación e interacción de la noción moderna de la naturaleza con nociones híbridas de cuasi-objetos y cuasi-humanos, así como la transformación de las concepciones de la naturaleza de una entidad apolítica a construcciones sociales con implicaciones políticas. Para analizar estos procesos, este texto presenta una revisión teórica de los cambios que se han dado en las categorías e investigaciones antropológicas sobre naturaleza, ecología y medio ambiente. Se destacan dos tendencias: la primera, ligada al replanteamiento de la dicotomía naturaleza/cultura a la luz de los conocimientos locales. La segunda, relacionada con los aportes de la ecología política, enfatizando el papel de los diferentes actores dentro de los discursos ambientales. Finalmente, se contextualizan las perspectivas de investigación sobre naturaleza, ecología y medio ambiente en América latina y en Colombia.

BWR fuel clad behaviour following LOCA

International Nuclear Information System (INIS)

Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

1996-01-01

Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

Naturaleza humana y política en Denis Diderot

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Pablo Scotto Benito

2015-02-01

Full Text Available El presente trabajo se propone relacionar la concepción que Diderot tiene del ser humano con sus ideas políticas. A pesar de que no se pueda hallar en la obra del “filósofo” un tratado sistemático en el que el estudio de la naturaleza humana sirva como fundamentación de un determinado sistema político, la epistemología empirista y el monismo materialista diderotianos, con la consecuente revalorización de las pasiones corporales, conducen a una moral y una política universalistas, acordes con la naturaleza humana y centradas en la felicidad de los individuos.

LOFT/L3-, Loss of Fluid Test, 7. NRC L3 Small Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the seventh in the NRC L3 Series of small-break LOCA experiments. A 2.5-cm (10-in.) cold-leg non-communicative-break LOCA was simulated. The experiment was conducted on 20 June 1980

Genetic diversity of wild germplasm of "yerba mate" (Ilex paraguariensis St. Hil.) from Uruguay.

Science.gov (United States)

Cascales, Jimena; Bracco, Mariana; Poggio, Lidia; Gottlieb, Alexandra Marina

2014-12-01

The "yerba mate" tree, Ilex paraguariensis St. Hil., is a crop native to subtropical South America, marketed for the elaboration of the highly popular "mate" beverage. The Uruguayan germplasm occupies the southernmost area of the species distribution range and carries adaptations to environments that considerably differ from the current production area. We characterized the genetic variability of the germplasm from this unexplored area by jointly analyzing individuals from the diversification center (ABP, Argentina, Brazil and Paraguay) with 19 nuclear and 11 plastidic microsatellite markers. For the Uruguayan germplasm, we registered 55 alleles (18 % private), and 80 genotypes (44 % exclusive), whereas 63 alleles (28.6 % private) and 81 genotypes (42 % exclusive) were recorded for individuals from ABP. Only two plastidic haplotypes were detected. Distance-based and multilocus genotype analyses showed that individuals from ABP intermingle and that the Uruguayan germplasm is differentiated in three gene-pools. Significant positive correlations between genetic and geographic distances were detected. Our results concur in that ABP individuals harbor greater genetic variation than those from the tail of the distribution, as to the number of alleles (1.15-fold), He (1.19-fold), Rs (1.39-fold), and the between-group genetic distances (1.16-fold). Also the shape of the genetic landscape interpolation analysis suggests that the genetic variation decays southward towards the Uruguayan territory. We showed that Uruguayan germplasm hosts a combination of nuclear alleles not present in the central region, constituting a valuable breeding resource. Future conservation efforts should concentrate in collecting numerous individuals of "yerba mate" per site to gather the existent variation.

Protective effect of yerba mate intake on the cardiovascular system: a post hoc analysis study in postmenopausal women

OpenAIRE

da Veiga, D.T.A.; Bringhenti, R.; Copes, R.; Tatsch, E.; Moresco, R.N.; Comim, F.V.; Premaor, M.O.

2018-01-01

The prevalence of cardiovascular and metabolic diseases is increased in postmenopausal women, which contributes to the burden of illnesses in this period of life. Yerba mate (Ilex paraguariensis) is a native bush from Southern South America. Its leaves are rich in phenolic components, which may have antioxidant, vasodilating, hypocholesterolemic, and hypoglycemic proprieties. This post hoc analysis of the case-control study nested in the Obesity and Bone Fracture Cohort evaluated the consumpt...

The large break LOCA evaluation method with the simplified statistic approach

International Nuclear Information System (INIS)

Kamata, Shinya; Kubo, Kazuo

2004-01-01

USNRC published the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology to large break LOCA which supported the revised rule for Emergency Core Cooling System performance in 1989. In USNRC regulatory guide 1.157, it is required that the peak cladding temperature (PCT) cannot exceed 2200deg F with high probability 95th percentile. In recent years, overseas countries have developed statistical methodology and best estimate code with the model which can provide more realistic simulation for the phenomena based on the CSAU evaluation methodology. In order to calculate PCT probability distribution by Monte Carlo trials, there are approaches such as the response surface technique using polynomials, the order statistics method, etc. For the purpose of performing rational statistic analysis, Mitsubishi Heavy Industries, LTD (MHI) tried to develop the statistic LOCA method using the best estimate LOCA code MCOBRA/TRAC and the simplified code HOTSPOT. HOTSPOT is a Monte Carlo heat conduction solver to evaluate the uncertainties of the significant fuel parameters at the PCT positions of the hot rod. The direct uncertainty sensitivity studies can be performed without the response surface because the Monte Carlo simulation for key parameters can be performed in short time using HOTSPOT. With regard to the parameter uncertainties, MHI established the treatment that the bounding conditions are given for LOCA boundary and plant initial conditions, the Monte Carlo simulation using HOTSPOT is applied to the significant fuel parameters. The paper describes the large break LOCA evaluation method with the simplified statistic approach and the results of the application of the method to the representative four-loop nuclear power plant. (author)

Canje de deuda externa por conservación de la naturaleza

Directory of Open Access Journals (Sweden)

Klaus Georg Binder

2001-12-01

Full Text Available El interés que tienen los países con grandes riquezas naturales en la reducción de su deuda externa y el interés que tienen los países industrializados en la conservación de la naturaleza se pueden combinar mediante un canje de deuda externa por conservación de naturaleza (Debt-for-Nature Swaps [swap (del inglés: cambio, canje]. Existen dos esquemas para el canje: (I Dos gobiernos, uno acreedor con interés en la protección de la naturaleza y otro deudor interesado en la disminución de la deuda externa, acuerdan la donación de la obligación a cambio de la conservación de la naturaleza. (II Un banco comercial de un país desarrollado, acreedor de un país en desarrollo con dificultades en la cancelación de la deuda, vende en el mercado secundario títulos de esta deuda por un precio menor que el de su valor nominal, pues sólo así logra disminuir la pérdida ocasionada por la no cancelación de dicha deuda. Una organización no gubernamental ambiental internacional compra la deuda y negocia con el país deudor la forma de cancelarla. Los canjes de deuda externa por conservación de la naturaleza representan un valioso complemento de la ayuda para el desarrollo orientada hacia el medio ambiente otorgada por los países industrializados, que sin duda deben desembolsar la contribución decisiva para la protección de ls grandes riquezas naturales en los países en desarrollo.

Hydrogen production in a PWR during LOCA

International Nuclear Information System (INIS)

Cassette, P.

1983-12-01

The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

Geografía Física y conservación de la Naturaleza

Directory of Open Access Journals (Sweden)

Francisco López Bermúdez

2002-01-01

Full Text Available El suelo, el agua, la vegetación y el paisaje, componentes básicos de la naturaleza, son recursos vitales y en gran parte no renovables que están sometidos a una presión humana cada vez mayor. Para que puedan desempeñar sus numerosas funciones, es necesario mantenerlos en buen estado y hacer de ellos una gestión y uso durables, y para ello, se precisa conocer las complejas y dinámicas relaciones que registran, en el marco del sistema naturaleza. La Geografía Física, como ciencia ambiental, socialmente útil, puede y debe contribuir al conocimiento de la funcionalidad y valores de la naturaleza y sus recursos, y a la relación de los humanos con ella.

Complete BWR--EM LOCA analysis using the WRAP--EM system

International Nuclear Information System (INIS)

Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

1979-01-01

The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

Parallel channel effects under BWR LOCA conditions

International Nuclear Information System (INIS)

Suzuki, H.; Hatamiya, S.; Murase, M.

1988-01-01

Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

Steam generator tube rupture effects on a LOCA

International Nuclear Information System (INIS)

LaChance, J.L.

1979-01-01

A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

LOFT/L3-6, Loss of Fluid Test, 6. NRC L3 Small Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the sixth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were running. The experiment was conducted on 10 December 1980

LOFT/L3-5, Loss of Fluid Test, 5. NRC L3 Small Break LOCA Experiment

International Nuclear Information System (INIS)

1991-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the fifth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were shut off. The experiment was conducted on 29 September 1980

Critical heat flux concerns during the flow instability phase of a DEGB LOCA

International Nuclear Information System (INIS)

Shadday, M.A. Jr.

1990-08-01

Arguments are presented that support the proposal that a separate burnout risk analysis, for the Flow Instability (FI) phase of a LOCA, not be required for reactor restart. With expected reactor power limits, flow instability will occur before critical heat flux (CHF). Since FI power limits preclude the occurrence of flow instability in a bounding accident, a DEGB LOCA, the risk of CHF and attendant burnout is negligible. A review of RDAP data revealed that in the past reactor assemblies operated at flow and power conditions similar to those expected in a LOCA without burnout occurring. This is strong bounding empirical evidence, without the scaling concerns of laboratory experiments. A bounding analysis of the influences of assembly non-idealities on CHF, power tilts, and channel eccentricity, is included. The margin between operating heat fluxes, during the postulated LOCA, and CHF was quantified by scoping calculations. Based on measured azimuthal power variations, the local heat flux would have to be more than 20 standard deviations above the calculated mean heat flux for CHF to occur

CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

International Nuclear Information System (INIS)

Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

2015-01-01

This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

Practical illustration of the traditional vers. alternative LOCA embrittlement criteria

International Nuclear Information System (INIS)

Vrtilkova, V.; Novotny, L.; Hamouz, V.; Doucha, R.; Tinka, I.; Macek, J.; Lahovsky, F.

2005-01-01

Evaluation of LOCA time behaviour is usually based on traditional embrittlement criterion, represented by the equivalent cladding reacted (ECR) limit 17 % (18 %) at the peak cladding temperature below 1204 0 C (1200 0 C). From different existing correlations for evaluation the ECR, the correlations of Baker-Just, Cathcart and VNIINM (Bibilashvili) are discussed here. Results, obtained by these correlations, are illustrated for typical and atypical LOCA courses analysed for the WWER 440 plant. An approach to assess these correlations from the viewpoint of violation of the observed criterion is presented. This approach is based on determination of the temperature vers. time of exposition, when the criterion limit is reached. Reasons leading to necessity of alternative criterion proposal are summarised. This criterion for LOCA events evaluation, including corresponding correlation, is proposed on the basis of the long-term experimental research of cladding materials at UJP Praha. The computational results, obtained according to this alternative criterion, are illustrated for the same courses of LOCA events as for traditional criteria and traditional correlations. Proposed criterion is also confronted with the other discussed criteria in accordance with mentioned approach presented in this paper. The characteristic experimental results and key findings are summarised. They substantiate and support the proposed alternative criterion. An advantage of the criterion is its independence on ECR, on hydrogen and oxygen content and on oxidation history, and its applicability to current Zr-based alloy cladding materials as well. This applicability is kept while preserving the simplicity of the criterion using. (author)

Naturaleza y libertad en el pensamiento de Simone de Beauvoir

Directory of Open Access Journals (Sweden)

Alicia H. Puleo

2009-01-01

Full Text Available Simone de Beauvoir es una de las filósofas que más ha contribuido a la transformación de las sociedades modernas con su reivindicación de la libertad de las mujeres. Este trabajo examina su visión de la Naturaleza y diferencia entre los aspectos de su legado que tienen un valor indiscutible para el feminismo contemporáneo y aquéllos que resultan problemáticos en una época de desarrollo de la ética ecológica y de revisión de los dualismos interconectados de Naturaleza y libertad, cuerpo y mente, sentimientos y razón.

A synopsis of experimental activities on small-break LOCA

International Nuclear Information System (INIS)

Hein, D.

1984-01-01

Through reactor safety studies like WASH 1400 or the ''Deutsche Risiko-Studie'' the attention has turned from large break loss of coolant accidents to small breaks because of the high contribution of this type of accidents to core meltdown. But only after the TMI-2 accident were also the main activities in the experimental fields shifted world-wide to the small break LOCAs. Since TMI numerous research programs have either been finished or are underway. This review paper presents: a classification of the various types of transients according to break size; a discussion of major physical phenomena associated with a small break LOCA, and a description of a few selected research programs and the most important results achieved. (author)

Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

International Nuclear Information System (INIS)

Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

1998-01-01

Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs

Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10

International Nuclear Information System (INIS)

1984-02-01

This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff-approved acceptance criteria for LOCA-related hydrodynamic loads are provided in Appendix C of this report

La ley de naturaleza como mandato divino. Continuidades entre los escritos tempranos y de madurez en la obra de John Locke

Directory of Open Access Journals (Sweden)

Joan Severo Chumbita

2015-01-01

Full Text Available El presente artculo tiene por objeto analizar el concepto de ley de naturaleza de John Locke. En este sentido, se establecern cinco rasgos fundamentales que se mantienen a lo largo de su obra. En primer lugar, el sesgo teolgico de la fundamentacin de la ley de naturaleza que resulta determinante respecto a las otras cuatro caractersticas. En oposicin a las interpretaciones en trminos profanos, se mostrar que para Locke la ley de naturaleza es siempre un mandato divino. En segundo lugar, se atender a su dimensin teleolgica, la cual se manifiesta en su concordancia con la naturaleza humana, sin la cual la observancia de la ley de naturaleza resultara imposible. En tercer lugar, se estudiar el carcter no-innato de la ley de naturaleza, vinculado directamente con la exterioridad al hombre que supone el hecho de que la ley de naturaleza sea un mandato divino, a pesar de adecuarse a la naturaleza humana para su realizacin. En cuarto lugar, se analizar la necesidad de interpretacin por parte de la razn individual. En efecto, la ley de naturaleza, en tanto mandato divino no-innato, exige una demostracin pues no hay de ella una captacin racional inmediata. Por ltimo, se abordar el lugar que ocupa la ley de naturaleza frente al derecho positivo. En este sentido, tanto el carcter externo, concordante con la naturaleza humana, no-innato e interpretable por parte de la razn individual, habilitan que la ley de naturaleza sirva de crtica a todo derecho positivo. De este modo, estos cinco elementos, articulados en el orden propuesto, permiten comprender el status epistemolgico y la funcin prctica del concepto de ley de naturaleza, fundamentales, como es sabido, a la concepcin lockeana de la propiedad, el Estado y la resistencia.

Simulation of fuel rod behaviour during various break LOCAs in PWRs

International Nuclear Information System (INIS)

Gadalla, A.A.; El-Fawal, M.M.

1996-01-01

During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

LA FARMACIA NATURALEZA – FUENTE DE FÁRMACOS EN EL SIGLO XXI

Directory of Open Access Journals (Sweden)

Oliver Callies

2011-12-01

Full Text Available La naturaleza ha sido fuente de fármacos desde el inicio de la historia y en muchas culturas lo sigue siendo hoy en día, como en la medicina tradicional china o la ayurveda. Por otra parte, un 50% de los fármacos actualmente usados en la medicina del mundo occidental derivan directa o indirectamente de la naturaleza, como la Aspirina® o el Taxol®. El camino que una sustancia recorre desde la naturaleza hasta convertirse en un nuevo fármaco es largo, laborioso y, en general, muy costoso. Los pasos más importantes a lo largo de este camino son el aislamiento de la fuente natural, elucidación estructural de los metabolitos aislados, sus ensayos en diferentes sistemas biológicos, hasta llegar a las fases clínicas y, finalmente - si todo va bien - la aprobación por la FDA o EMEA. La aparición de nuevas enfermedades infecciosas, la persistencia de enfermedades que aún no tienen tratamiento o la aparición de resistencia, amenazan la eficacia de los medicamentos actualmente en uso. La naturaleza nos puede proporcionar nuevas sustancias más eficaces y quizá con menos efectos secundarios para enfrentarnos a las enfermedades más impactantes de nuestro tiempo y del futuro.

Development and Assessment of the Appendix K Version of RELAP5-3D for LOCA Licensing Analysis

International Nuclear Information System (INIS)

Liang, Thomas K.S.; Chang, C.-J.; Hung, H.-J.

2002-01-01

In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss-of-coolant accident (LOCA) would limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during a LOCA, it generally takes much more resources to develop. Instead, implementation of evaluation models required by Appendix K of 10CFR50 on an advanced thermal-hydraulic platform such as RELAP5, TRAC, etc., also can gain significant margin for the PCT calculation. Through compliance evaluation against Appendix K of 10CFR50, all of the required evaluation models have been implemented in RELAP5-3D. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effects experiments and eight sets of LOCA integral experiments were adopted. Through the assessments against separate-effects experiments, the success of the code modification in accordance with Appendix K of 10CFR50 was demonstrated. Besides, one set of a typical integral large-break LOCA from Loss-of-Fluid Test Facility experiments (L2-5) has also been applied to preliminarily evaluate the integral performance of the Appendix K version of RELAP5-3D. The PCT predicted by the evaluation models is greater than the one from best-estimate calculation in the whole LOCA history with the conservatism of 150 K, and the measured PCTs of L2-5 are also well bounded by the evaluation model calculation. Another seven sets of integral-effect experiments will be further applied in the next step to ensure the reasonable integral conservatism of the newly developed LOCA licensing analysis code (RELAP5-3DK/INER), which can cover all the phases of both large- and small LOCA in one code

Bulging of pressure tubes at hot spots under LOCA conditions

International Nuclear Information System (INIS)

Manu, C.; Shewfelt, R.S.W.; Wright, A.C.D.; Aboud, R.; Lau, J.H.K.; Sanderson, D.B.

1996-01-01

During certain postulated loss-of-coolant accidents (LOCA) in a CANDU reactor, some fuel channels can become highly voided within a very short time. Although the pressure tubes are heated mainly by convection and thermal radiation during the LOCA transient, additional heat flow occurs through the bearing pads that are in contact with the pressure tribe. This contact can lead to local hot spots and associated thermal stresses in the pressure tube wall. The two factors that affects the behavior of the pressure tubes during LOCA conditions are the internal pressure and the local heating. Although the effect of internal pressure and of axially uniform temperature has been studied elsewhere, the effect of the local heating on the pressure tube behavior has not been modelled before. This paper shows that the bulging of a pressure tube at a hot spot is the result of the thermal stresses that are developed in a pressure tube during a LOCA transient. To isolate the local heating effect from the internal pressure, a series of single-effect experiments was performed. In these experiments, sections of a CANDU pressure tube were subjected to local heating only. The thermal profile and the local deformation were measured function of time. To quantify the effect of the thermal stresses on the bulging of pressure tubes at hot spots and to develop numerical tools that can predict such bulging, finite element analyses were performed rising the ABAQUS finite element computer code. Use of the measured thermal profiles in the ABAQUS finite element analysis, resulted in very good agreement between the predicted and measured displacements. (author)

Unas opciones para muestras de la naturaleza

Science.gov (United States)

Frank H. Wadsworth

2009-01-01

La ciencia, incluso la de Puerto Rico, aprende de muestras, su interpretación y extrapolación a situaciones más extensas. Las muestras de la naturaleza de Puerto Rico: las del suelo, la hojarasca, la sombra, la biomasa, los árboles, las aves, los reptiles y anfibios, o los insectos caracterizan los ecosistemas. [article in Spanish

Modelling of LOCA Tests with the BISON Fuel Performance Code

Energy Technology Data Exchange (ETDEWEB)

Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

Development of Advanced Non-LOCA Analysis Methodology for Licensing

International Nuclear Information System (INIS)

Jang, Chansu; Um, Kilsup; Choi, Jaedon

2008-01-01

KNF is developing a new design methodology on the Non-LOCA analysis for the licensing purpose. The code chosen is the best-estimate transient analysis code RETRAN and the OPR1000 is aimed as a target plant. For this purpose, KNF prepared a simple nodal scheme appropriate to the licensing analyses and developed the designer-friendly analysis tool ASSIST (Automatic Steady-State Initialization and Safety analysis Tool). To check the validity of the newly developed methodology, the single CEA withdrawal and the locked rotor accidents are analyzed by using a new methodology and are compared with current design results. Comparison results show a good agreement and it is concluded that the new design methodology can be applied to the licensing calculations for OPR1000 Non-LOCA

Operator reliability analysis during NPP small break LOCA

International Nuclear Information System (INIS)

Zhang Jiong; Chen Shenglin

1990-01-01

To assess the human factor characteristic of a NPP main control room (MCR) design, the MCR operator reliability during a small break LOCA is analyzed, and some approaches for improving the MCR operator reliability are proposed based on the analyzing results

SCRELA, LOCA Analysis of Super-Critical Light-Water Reactors

International Nuclear Information System (INIS)

Lee, J.H.; Koshizuka, S.; Oka, Y.

2001-01-01

Description of program or function: LOCA Analysis Code for the Supercritical-Water Cooled Reactor. - Blowdown Module: Calculation of the Blowdown Phase and Refill Phase. - Reflood Module: Calculation of the Reflood Phase

Cuadros y escrituras de la naturaleza

Directory of Open Access Journals (Sweden)

Pimentel, Juan

2004-12-01

Full Text Available To conmemorate Humboldt return journey to Europe means that of a double transference: we also celebrate how the New World was back to the Old, and how Nature back to the Book. Our focus is representation, an sphere where humboldtian descriptive literature plays a basic role. Here we study the relation between images and words, between Humboldt visual thinking and his narrative techniques in order to animate, giving life to nature. Views of Nature and Cosmos are the two main works in this field; Bernardin de Saint-Pierre, Goethe, and Aristotle, some of the keys we use to decode that descriptive literature.

El aniversario del regreso de Humboldt a Europa es también el de una doble transferencia: el del Nuevo al Viejo Mundo y el de la Naturaleza al libro. Este trabajo se centra en la representación, esfera en la que la literatura descriptiva humboldtiana desempeña una labor fundamental. Estudiamos la relación entre la imagen y la palabra, entre el pensamiento visual de Humboldt y sus técnicas narrativas para animar y vivificar la naturaleza. Cuadros y Cosmos son las obras más destacadas en este terreno; Bernardin de Saint-Pierre, Goethe y Aristóteles, algunas de las claves que empleamos para descifrar dicha literatura descriptiva.

Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

International Nuclear Information System (INIS)

Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

1987-01-01

The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Suh, Jeong Kwan, E-mail: jksuh@khnp.co.kr [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kim, Jae Won; Kwon, Sun Guk; Lee, Jae Yong [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Cho, Hyoung Kyu; Park, Goon Cherl [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

2015-08-15

Highlights: • In-vessel downstream effect tests were performed in the presence of LOCA-generated debris. • Available driving heads under each LOCA scenario were verified using experimental data. • Fibrous debris was prepared to satisfy the length distribution obtained from the bypass test. • Limiting test conditions were identified through sensitivity studies. - Abstract: Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

Analysis of a convection loop for GFR post-LOCA decay heat removal

International Nuclear Information System (INIS)

Williams, W.C.; Hejzlar, P.; Saha, P.

2004-01-01

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA gas-cooled fast reactor (GFR). The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO 2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO 2 outdoes helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops. (authors)

PWR cold-leg small break loca with faulty HPI

International Nuclear Information System (INIS)

Kumamaru, H.; Kukita, Y.

1991-01-01

The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

Energy Technology Data Exchange (ETDEWEB)

Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

2015-10-15

Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of

Sensitivity Study on Analysis of Reactor Containment Response to LOCA

International Nuclear Information System (INIS)

Chung, Ku Young; Sung, Key Yong

2010-01-01

As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

Sensitivity Study on Analysis of Reactor Containment Response to LOCA

Energy Technology Data Exchange (ETDEWEB)

Chung, Ku Young; Sung, Key Yong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2010-10-15

As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

Safety assessment of the advanced CANDU reactor in postulated LOCA/LOECC events

International Nuclear Information System (INIS)

Hazen Hezhi Fan; Zoran Bilanovic

2005-01-01

The Advanced CANDU Reactor TM (ACR TM ) retains the proven strengths and features of CANDU reactors, and incorporates innovative new features and state-of-the-art technology. In addition to the enhanced emergency core cooling system, the reserve water system is designed to be available to inject reserve water by gravity into the reactor inlet headers after a postulated loss-of-coolant accident (LOCA). To assist in the ACR design and analysis of beyond the design basis events, simulations are needed to demonstrate the effectiveness of these two independent systems on core cooling, and to assess the consequences of the postulated accident coincident with the impairment of either of the two systems. The current paper is subject to an assessment of a postulated large LOCA coincident with loss of the emergency core cooling (LOECC) system. A postulated LOCA/LOECC has very low probability, in the range usually associated with severe core damage events. However, in the CANDU design, including ACR, the presence of moderator water surrounding the fuel channels acts as an effective heat sink, together with other safety features, to prevents severe core damage following a postulated LOCA/LOECC. Therefore, it is possible to analyse LOCA/LOECC using the same deterministic tools that are used for analysis of events with much higher frequencies, in the design basis event range. The assessment is conducted based on the current ACR-700 design. However, the analysis methodology, scope, computer tools, and the results in principle, are applicable to larger ACR designs. This assessment includes system (circuit), fuel channel, and fuel analyses. Some assessment results are needed in subsequent moderator analysis and containment analysis. In the assessment, several simulations were performed to analyse the full circuit and individual fuel channel transient behaviours, as well as the fission product release behaviour. The assessment has captured the key responses of the reactor heat

Educación ambiental, inteligencia espiritual y naturaleza

Directory of Open Access Journals (Sweden)

Jordi PUIG BAGUER

2014-12-01

Full Text Available La conciencia ambiental es característica en nuestra cultura, y plantea múltiples retos educativos. Las elecciones ambientales no dejan de construirnos… o destruirnos, tanto personal como colectivamente, en dimensiones corporales o culturales, en el individuo humano y en su paisaje. Entre ser humano y naturaleza existe un vínculo, común a ambos, de humanidad y naturalidad que no puede olvidarse en la educación sin erosionar el valor y el respeto del ser humano y la naturaleza a la vez.Este artículo busca impulsar la educación en medio ambiente mediante el desarrollo de la inteligencia espiritual y viceversa, con el fin de alimentar nuestro conocimiento con una profundidad y universalidad que necesitamos. En particular, destaca el papel cultural de la ciencia ecológica y resalta que la diferencia esencial entre los seres humanos y su medio natural permite –y debe– alimentar el respeto de ambos a través de la importancia y valor del recíproco. 

Buen vivir con la naturaleza en las instituciones educativas: una necesidad en Boyacá, Colombia

Directory of Open Access Journals (Sweden)

Aracely Burgos Ayala

2016-01-01

Full Text Available El actual modelo de desarrollo económico ha logrado desequilibrar la relación entre los humanos y los restantes seres de la naturaleza. Su mitigación es posible a través del desarrollo de prácticas del buen vivir ( bv en dos de sus variables y dimensiones derivadas: los derechos humanos (educación y participación y los derechos de la naturaleza (respeto, protección y restauración en las instituciones educativas ( ie . Estos son escenarios de impacto social que posibilitan nuevas generaciones de ciudadanos hacia una vida humana y ecológicamente más justa; el desafío es construir esta sociedad con educación. Con tal propósito, se realizan aproximaciones teóricas hacia: 1 comprender el bv con la naturaleza; 2 articular variables y dimensiones del bv en las ie , y 3 presentar la necesidad de indagar por el estado actual del bv con la naturaleza en las ie de Boyacá, Colombia.

A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

International Nuclear Information System (INIS)

Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

2015-01-01

Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Ho [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of); Chang, Soon Heung [Handong Global University, 558, Handong-ro, Buk-gu, Pohang Gyeongbuk 37554 (Korea, Republic of); Choi, Yu Jung [Korea Hydro and Nuclear Power Co.—Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 34101 (Korea, Republic of); Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of)

2015-12-15

Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

International Nuclear Information System (INIS)

Casal, V.; Malang, S.; Rust, K.

1982-10-01

For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

LOFT/L2-3, Loss of Fluid Test, 2. NRC L2 Large Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the second of the NRC L2 Series of nuclear large Break LOCA experiments, and was conducted on 12 May 1979. It simulated a 100% cold leg break with a maximum heat generation of 39 kW/m

Statistics and integral experiments in the verification of LOCA calculations models

International Nuclear Information System (INIS)

Margolis, S.G.

1978-01-01

The LOCA (loss of coolant accident) is a hypothesized, low-probability accident used as a licensing basis for nuclear power plants. Computer codes which have been under development for at least a decade have been the principal tools used to assess the consequences of the hypothesized LOCA. Models exist in two versions. In EM's (Evaluation Models) the basic engineering calculations are constrained by a detailed set of assumptions spelled out in the Code of Federal Regulations (10 CFR 50, Appendix K). In BE Models (Best Estimate Models) the calculations are based on fundamental physical laws and available empirical correlations. Evaluation models are intended to have a pessimistic bias; Best Estimate Models are intended to be unbiased. Because evaluation models play a key role in reactor licensing, they must be conservative. A long-sought objective has been to assess this conservatism by combining Best Estimate Models with statisticallly established error bounds, based on experiment. Within the last few years, an extensive international program of LOCA experiments has been established to provide the needed data. This program has already produced millions of measurements of temperature, density, and flow and millions of more measurements are yet to come

Identification of Error of Commissions in the LOCA Using the CESA Method

Energy Technology Data Exchange (ETDEWEB)

Tukhbyet-olla, Myeruyert; Kang, Sunkoo; Kim, Jonghyun [KEPCO international nuclear graduate school, Ulsan (Korea, Republic of)

2015-10-15

An Errors of commission (EOCs) can be defined as the performance of any inappropriate action that aggravates the situation. The primary focus in current PSA is placed on those sequences of hardware failures and/or EOOs that lead to unsafe system states. Although EOCs can be treated when identified, a systematic and comprehensive treatment of EOC opportunities remains outside the scope of PSAs. However, some past experiences in the nuclear industry show that EOCs have contributed to severe accidents. Some recent and emerging human reliability analysis (HRA) methods suggest approaches to identify and quantify EOCs, such as ATHEANA, MERMOS, GRS, MDTA, and CESA. The CESA method, developed by the Risk and Human Reliability Group at the Paul Scherrer Institute, is to identify potentially risk-significant EOCs, given an existing PSA. The main idea underlying the method is to catalog the key actions that are required in the procedural response to plant events and to identify specific scenarios in which these candidate actions could erroneously appear to be required. This paper aims at identifying EOCs in the LOCA by using the CESA method. This study is focused on the identification of EOCs, while the quantification of EOCs is out of scope. Then, this paper applies the CESA method to the emergency operating procedure (EOP) of LOCA for APR1400. Finally, this study presents potential EOCs that may lead to the aggravation in the mitigation of LOCA. This study has identified the EOC events for APR1400 in the LOCA using CESA method. The result identified three candidate EOCs event using operator action catalog and RAW cutset of LOCA. These candidate EOC events are inappropriate terminations of safety injection system, safety injection tank and containment spray system. Then after reviewing top 100 accident sequences of PSA, this study finally identified one EOC scenario and EOC path, that is, inappropriate termination of safety injection system.

Best estimate LB LOCA approach based on advanced thermal-hydraulic codes

International Nuclear Information System (INIS)

Sauvage, J.Y.; Gandrille, J.L.; Gaurrand, M.; Rochwerger, D.; Thibaudeau, J.; Viloteau, E.

2004-01-01

Improvements achieved in thermal-hydraulics with development of Best Estimate computer codes, have led number of Safety Authorities to preconize realistic analyses instead of conservative calculations. The potentiality of a Best Estimate approach for the analysis of LOCAs urged FRAMATOME to early enter into the development with CEA and EDF of the 2nd generation code CATHARE, then of a LBLOCA BE methodology with BWNT following the Code Scaling Applicability and Uncertainty (CSAU) proceeding. CATHARE and TRAC are the basic tools for LOCA studies which will be performed by FRAMATOME according to either a deterministic better estimate (dbe) methodology or a Statistical Best Estimate (SBE) methodology. (author)

Contenido y aplicación de los derechos de la naturaleza

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René Bedón Garzón

2016-12-01

Full Text Available La Constitución de la República del Ecuador de 2008 ha consagrado derechos a favor de la naturaleza, incluyendo una reserva constitucional para su creación. Durante estos años estos derechos han sido mejor definidos por la legislación y por la jurisprudencia, según se muestra en la presente investigación. Desde el año 2008 se han presentado varias acciones de protección y medidas cautelares constitucionales a fin de hacer efectivos estos derechos. Tales procesos han terminado con decisiones jurisdiccionales que, buscando garantizar el derecho de la naturaleza a la conservación integral y los derechos de las comunidades afectadas, han dictaminado la suspensión de obras y la obtención de los permisos ambientales correspondientes por parte del Estado para no generar impactos ambientales. Además han aplicado el principio precautorio, han suspendido actividades aunque no haya evidencia científica de daño, y se ha ponderado derechos para permitir la limitación del derecho a la propiedad privada, a fin de que se realicen tareas de remediación de un evento ambiental y se logre garantizar el derecho de la naturaleza a la restauración.

Proceedings of the joint CSNI/CNRA workshop on redefining the large break LOCA: technical basis and its implications

International Nuclear Information System (INIS)

2003-01-01

The objective of the Workshop was to facilitate an exchange of information on a topic, which could potentially impact both the operation of current reactors and the design of future reactors. A number of OECD countries were actively working in this area at the moment. Regulators, Researchers and Industry representatives needed to exchange information on the current regulation and technical issues associated with the Large Break LOCA (LB-LOCA), and to further discuss rationales and motives which could lead to a redefinition of the LB- LOCA. The focus was on design and safety implications. Policy issues were not discussed but the workshop provided technical inputs for policy makers. The workshop covered different reactor designs (CANDUs, VVERs, LWRs). The workshop was articulated over three questions: 1. What drives the need to redefine the LB-LOCA? There was a consensus on well founded drivers for the redefinition and on observations that, if the change is made right, it can enhance the safety of nuclear power plants and also reduce the costs of power production. Three papers were presented. In the first paper, Mr Bajorek (USNRC) discussed the redefinition of the LB-LOCA in the context of risk informing their regulations. Mr. Bajorek emphasised that redefining LB-LOCA is being considered by US NRC from a risk perspective to improve the safety focus and that regulators should better focus on safety and risk contributors and thereby formulate the regulations to better use available resources. He added that the present LOCA definition has not only a great impact on the plant design but also on operating limits according to the Technical Specifications as well as on testing conditions. In the second paper, Mr Pietrangelo (NEI, USA) presented a paper on the need to redefine the large break LOCA from the industrial viewpoints. He emphasised that the strong leadership commitments by both NRC and industry are necessary, and that redefining LB- LOCA is central to risk

Simulated LOCA Test and Characterization Study Related to High Burn-Up Issue

International Nuclear Information System (INIS)

Park, D. J.; Jung, Y. I.; Choi, B. K.; Park, S. Y.; Kim, H. G.; Park, J. Y.

2012-01-01

For the safety evaluation of fuel cladding during the injection of emergency core coolant, simulated Loss-of-coolant accident (LOCA) test was performed by using Zircaloy-4 fuel cladding samples. Zircaloy-4 tube samples with and without prehydring were oxidized in a steam environment with the test temperature of 1200 .deg. C. Prehydrided cladding was prepared from as-fabricated Zircaloy-4 to study the effects of hydrogen on mechanical properties of cladding during high temperature oxidation and quench conditions. In order to measure the ductility of the tube samples embrittled by quenching water, ring compression test was carried out by using 8 mm ring sample sectioned from oxidized tube sample and microstructural analysis was also performed after simulated LOCA test. The results showed that hydrogen increases oxygen solubility and pickup rate in the beta layer. This reduces ductility of prehydrided fuel cladding compared with as-fabricated cladding. Trend in ductility decrease for prehydrided sample under simulated LOCA condition was very similar with data obtained from tests conducted using irradiated high burn-up fuel claddings

Data management system for full core LOCA-analysis using TRANSURANUS

International Nuclear Information System (INIS)

Maertens, D.; Spykman, G.

2005-01-01

A data management system has been developed to perform full core pin by pin calculations of normal operation and (LOCA-) transient behaviour of fuel rods. The system automatically generates the input from a data base, controls the fuel rod calculations and provides a powerful tool for visualising the results. The full core pin by pin analysis now allows to use specific power histories, rod geometries and material data as well as enveloping data. Fuel rod code Transuranus is used for the normal operation and the transient phase in one run, thus assuring that the calculated rod properties of the normal operation (pre-transient) phase are handed over in all detail and not compressed to the transient phase. Transuranus has been upgraded with respect to high temperature models for Zry and M5 TM -cladding for creep, oxidation, heat rate dependent phase transition and anisotropy in the α and the mixed crystal phase. Parameter studies have been carried out to investigate the influence of using rod specific power histories instead of enveloping power histories in a full core analysis. The results show a significant increase in the ratio of failed fuel rods during a LOCA transient from 0.12% to approx. 50%. Another study for a typical PWR LOCA transient shows very good correlation between the distribution of failed fuel rods and rods with significant ballooning. (author)

Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10. Final report

International Nuclear Information System (INIS)

Fields, M.B.; Kudrick, J.A.

1984-08-01

This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases and conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff approved acceptance criteria for LOCA-related hydrodynamic loads are provided in an appendix

La calidad y salud del suelo influyen sobre la naturaleza y la sociedad

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Hernán Burbano Orjuela

2017-02-01

Full Text Available La preservación del suelo resulta prioritaria para todos los sectores de la sociedad, porque su deterioro repercute adversamente sobre otros componentes de la naturaleza y sobre los grupos sociales que necesitan del suelo. Por eso, es importante mantener la calidad y salud del suelo, para que la sociedad pueda garantizar su seguridad alimentaria y disfrutar de los servicios ecosistémicos que presta este recurso natural. En este artículo se revisan los conceptos centrales que buscan explicar las relaciones del suelo con la naturaleza y con las personas.

Korean Consortium's preliminary research for enhancing a probabilistic fracture mechanics code, PRO-LOCA

International Nuclear Information System (INIS)

Kim, Sun-Hye; Park, Jung-Soon; Lee, Jin-Ho; Yun, Eun-Sub; Kang, Sun-Ye; Shim, Do-Jun

2015-01-01

The Battelle developed a probabilistic fracture mechanics code called PRO-LOCA, which can be used as a tool for evaluating the pipe break frequency. It is being further developed through the international co-operative research program, PARTRIDGE. KINS, KHNP-CRI, and KEPCO-E&C are participating in the PARTIRDGE program by composing a Korean Consortium. The members of Korean Consortium performed benchmark analyses using the beta version of PRO-LOCA 4.0 to evaluate the effect of variables such as simulation methods, crack features, loading conditions, and inspection models on the failure probabilities. The benchmark analyses showed that the PRO-LOCA can provide a trend consistent with the expected crack growth and pipe failure behavior. Especially, the availability of the stress intensity factor and crack opening displacement for non-idealized through-wall cracks was proven from this study. This new solution for non-idealized through-wall cracks had been developed by the Korean Consortium and it was newly included in PRO-LOCA 4.0. However, further improvement is needed to address the problems such as the instability of adaptive sampling method and the unexpected trend of failure probabilities at the early stage of crack growth

Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

International Nuclear Information System (INIS)

Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

1996-01-01

Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

LOCA analysis of SCWR-M with passive safety system

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

2013-06-15

Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

LOFT/L2-5, Loss of Fluid Test, 3. NRC L2 Large Break LOCA Experiment

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the third of the NRC L2 Series of nuclear large Break LOCA experiments, conducted on 16 June 1981. It simulated a 100% cold leg break with a maximum heat generation of 40 kW/m and rapid pump coast down

Modeling study of droplet behavior during blowdown period of large break LOCA based on experimental data

International Nuclear Information System (INIS)

Sakaba, Hiroshi; Umezawa, Shigemitsu; Teramae, Tetsuya; Furukawa, Yuji

2004-01-01

During LOCA (Loss Of Coolant Accident) in PWR, droplets behavior during blowdown period is one of the important phenomena. For example, the spattering from falling liquid film that flows from upper plenum generates those droplets in core region. The behavior of droplets in such flow has strong effect for cladding temperature behavior because these droplets are able to remove heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface. For safety analysis of LOCA in PWR, it is necessary to evaluate droplet diameter precisely in order to predict fuel cladding temperature changing by the calculation code. Based on the test results, a new droplet behavior model was developed for the MCOBRA/TRC code that predicts the droplet behavior during such LOCA events. Furthermore, the verification calculations that simulated some blowdown tests were performed using by the MCOBRA/TRAC code. These results indicated the validity of this droplet model during blow down cooling period. The experiment was focused on investigating the Weber number of steady droplet in the blow down phenomenon of large break LOCA. (author)

The deformation of PWR fuel in a LOCA

International Nuclear Information System (INIS)

Mann, C.A.; Hindle, E.D.; Parsons, P.D.

1982-04-01

Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 1500 0 C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

2018-01-01

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

Evaluation of Gap Conductance Approach for Mid-Burnup Fuel LOCA Analysis

International Nuclear Information System (INIS)

Lee, Joosuk; Woo, Swengwoong

2013-01-01

In this study, therefore, the applicability of gap conductance approach on the mid-burnup fuel in LOCA analysis was estimated in terms of the comparison of PCT distribution method means the fuel rod uncertainty is taken into account by the combination of overall uncertainty parameters of fuel rod altogether by use of a simple random sampling(SRS) technique. There are many uncertainty parameters of fuel rod that can change the PCT during LOCA analysis, and these have been identified by the authors' previous work already. But, for the 'best-estimate' LOCA safety analysis the methodology that dose not use the overall uncertainty parameters altogether but used the gap conductance uncertainty alone has been developed to simulate the overall fuel rod uncertainty, because it can represent many uncertainty parameters. Based on this approach, uncertainty range of gap conductance was prescribed as 0.67∼1.5 in audit calculation methodology on LBLOCA analysis. This uncertainty was derived from experimental data of fresh or low burnup fuel. Meanwhile, recent research work identify that the currently utilized uncertainty range seems to be not enough to encompass the uncertainty of mid-burnup fuel. Instead it has to be changed to 0.5∼2.4 for the mid-burnup fuel(30 MWd/kgU)

Evaluation of Gap Conductance Approach for Mid-Burnup Fuel LOCA Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In this study, therefore, the applicability of gap conductance approach on the mid-burnup fuel in LOCA analysis was estimated in terms of the comparison of PCT distribution method means the fuel rod uncertainty is taken into account by the combination of overall uncertainty parameters of fuel rod altogether by use of a simple random sampling(SRS) technique. There are many uncertainty parameters of fuel rod that can change the PCT during LOCA analysis, and these have been identified by the authors' previous work already. But, for the 'best-estimate' LOCA safety analysis the methodology that dose not use the overall uncertainty parameters altogether but used the gap conductance uncertainty alone has been developed to simulate the overall fuel rod uncertainty, because it can represent many uncertainty parameters. Based on this approach, uncertainty range of gap conductance was prescribed as 0.67∼1.5 in audit calculation methodology on LBLOCA analysis. This uncertainty was derived from experimental data of fresh or low burnup fuel. Meanwhile, recent research work identify that the currently utilized uncertainty range seems to be not enough to encompass the uncertainty of mid-burnup fuel. Instead it has to be changed to 0.5∼2.4 for the mid-burnup fuel(30 MWd/kgU)

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

Energy Technology Data Exchange (ETDEWEB)

Fleming, K.N., E-mail: KarlFleming@comcast.net [KNF Consulting LLC, Spokane, WA (United States); Lydell, B.O.Y. [SIGMA-PHASE INC., Vail, AZ (United States)

2016-08-15

Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

International Nuclear Information System (INIS)

Fleming, K.N.; Lydell, B.O.Y.

2016-01-01

Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

LOCA simulation tests in the RD-12 loop with multiple heat channels

International Nuclear Information System (INIS)

Ardron, K.H.; McGee, G.R.; Hawley, E.H.

1985-11-01

A series of tests has been performed in the RD-12 loop to study the bahaviour of a CANDU-type, primary heat transport system (PHTS) during the blowdown and injection phases of a loss-of-coolant accident (LOCA). Specifically, the tests were used to investigate flow stagnation and refilling of the core following a LOCA. RD-12 is a pressurized water loop with the basic geometry of a CANDU reactor PHTS, but at approximately 1/125 volume scale. The loop consists of U-tube steam generators, pumps, headers, feeders, and heated channels arranged in the symmetrical figure-of-eight configuration of the CANDU PHTS. In the LOCA simulation tests, the loop contained four horizontal heated channels, each containing a seven-element assembly of indirectly heated, fuel-rod simulators. The channels were nominally identical, and were arranged in parallel pairs between the headers in each half-circuit. Tests were carried out using various restricting orifices to represent pipe breaks of different sizes. The break sizes were specifically chosen such that stagnation conditions in the heated channels would be likely to occur. In some tests, the primary pumps were programmed to run down over a 100-s period to simulate a LOCA with simultaneous loss of pump power. Test results showed that, for certain break sizes, periods of low flow occurred in the channels in one half of the loop, leading to flow stratification and sheath temperature excursions. This report reviews the results of two of the tests, and discusses possible mechanisms that may have led to the low channel flow conditions observed in some cases. Plans for future experiments in the larger scale RD-14 facility are outlined. 5 refs

Cuerpo y naturaleza humana en la obra de Hannah Arendt

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Nicolás Patierno

2016-06-01

Full Text Available http://dx.doi.org/10.5007/1807-1384.2016v13n2p1 Comúnmente suele asociarse a Hannah Arendt con temas como la filosofía política, los totalitarismos y una postura crítica hacia los pilares de la modernidad, heredada de su vinculación teórica con Jaspers y Heidegger. Pese a esa lectura tradicional, la obra de la autora abarca diversas cuestiones histórico-políticas emparentadas con las nociones de cuerpo y naturaleza; entre otras problemáticas de interés social. Asumiendo una metodología histórica-hermenéutica basada en la revisión documental de fuentes primarias y secundarias, el objetivo del presente artículo radica en el rastreo de dichos conceptos a través de sus principales obras. Entre éstas se encuentran Los orígenes del totalitarismo y La condición humana como escritos fundamentales a la hora de interpretar el concepto de cuerpo subordinado a la esfera política. Arendt desmiente la existencia de una naturaleza humana innata, prescrita e incuestionable. Es a partir de la experiencia de los campos de concentración, que la idea de naturaleza, de origen “natural” de los pueblos, quedó relegada a la interpretación e instrumentación en manos del orden político imperante, concluyendo que el totalitarismo, es el mayor exponente de esta pretensión absoluta de apropiación de “lo natural”.

Core heatup prediction during SB LOCA with RELAP5/MOD3.2.2 Gamma

International Nuclear Information System (INIS)

Parzer, I.; Mavko, B.; Petelin, S.

2001-01-01

The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and the ability of RELAP5/MOD3.2.2 Gamma to predict core overheating. The code prediction has been compared to the three experiments, one conducted on the separate effect test facility NEPTUN in Switzerland and the other two conducted on two integral test facilities, PMK-2 in Hungary and PACTEL facility in Finland. In the case of a series of boiloff experiments performed on the NEPTUN test facility the influence of the two correlations available in MOD3.2.2 Gamma for determining interphase drag has been studied. In the case of IAEA-SPE-4 experiment simulation on PMK-2 facility the main goal of the analysis was to study the adequate modeling of the hexagonal core channel with 19-rod bundle and the phenomena during the core uncovering. The third analyzed experiment, OECD-ISP-33, was performed on PACTEL facility to study different natural circulation modes during SB LOCA. The analysis also focused on the final stage of this SB LOCA experiment, when core dryout and heatup was observed due to gradual emptying of the primary system. Following the experience the appropriate modeling options have been used to achieve better representation of the important phenomena during the SB LOCA.(author)

LOCA scenario tests of irradiated fuel rod specimens

International Nuclear Information System (INIS)

Scott, Harold

2004-01-01

Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

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Cesare Frepoli

2008-01-01

Full Text Available Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

Mechanical interaction between fuel pins and assemblies during LOCA in BWR

International Nuclear Information System (INIS)

Jonsson, T.

1978-10-01

The size of the rod elongation by oxidation is so large that deformation of a standard BWR fuel element with tie rods in the outer row will surely occur during a LOCA transient typical for BWRs with external pumps. Available data does not however show whether this deformation will occur early in the transient or during the cooling. Combined effects of thermal expansion of zircaloy and expansion due to oxidation and dissolution of oxygen can be expected to be large enough to cause rod bowing early in a LOCA transient. It is however not impossible that observed residual expansion of zircaloy tubes to a dominating extent are caused through expansion of zirconium oxide during cool-down. Length measurements of zircaloy tubes during a transient are desirable. (author)

Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

International Nuclear Information System (INIS)

Arshad, M. W.

2012-01-01

Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

About criteria of inadmissible embrittlement of zirconium fuel cladding during LOCA in the PWRs

International Nuclear Information System (INIS)

Osmachkin, V.S.

1999-01-01

According the licensing procedures the designers of the PWRs reactor have to prove the meeting of special safety requirements. One criteria on effectiveness of the Emergency Core Cooling System is not to exceeding some limited conditions of the fuel cladding during LOCA accidents (typical example T m ax o C, ECR<0,17 and oth.). The damage of fuel element in the core during LOCA is caused by the oxidation of the cladding, its embrittlement and thermal shock stresses after initiation of the heat removal by a cold water from emergency core cooling system. In the paper the conservatism in criteria to avoid brittle ruptures of the fuel elements is discussed. Taking into account the influence of fuel burnup on the property of the cladding and a potential presence of air in the steam, it is believed that criteria of survivability of the zircaloy fuel cladding during LOCA may not be enough conservative.(author)

Application of thermal hydraulic and severe accident code SOCRAT/V3 to bottom water reflood experiment QUENCH-LOCA-0

International Nuclear Information System (INIS)

Vasiliev, A.D.; Stuckert, J.

2013-01-01

Highlights: ► QLOCA-0 test simulates a design basis LOCA NPP accident with maximum temperature 1300 K. ► Deep understanding of hydraulics and thermal mechanics under accident conditions is necessary. ► We model the test QLOCA-0 with bottom flooding using the Russian code SOCRAT/V3. ► Calculated and experimental data are in a good agreement. ► Experimental procedure is determined to reach a representative LOCA scenario in future tests. -- Abstract: The thermal hydraulic and SFD (severe fuel damage) best estimate computer modeling code SOCRAT/V3 has been used for the calculation of QUENCH-LOCA-0 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (loss of coolant accident) nuclear power plant accident sequence in which the overheated up to 1300 K reactor core would be reflooded from the bottom by ECCS (emergency core cooling system). The first test QUENCH-LOCA-0 was successfully conducted at the KIT, Karlsruhe, Germany, in July 22, 2010, and was performed as the commissioning test for this series. The rod claddings are identical to that used in PWRs. The bundle was electrically heated in steam from 800 K to 1340 K with the heat-up rate of approximately 2.7 K/s. After cooling in the saturated steam the bottom flooding with water flow rate of about 100 g/s was initiated. The SOCRAT calculated results are in a good agreement with experimental data taking into account additional quenching due to water condensate entrainment at the steam cooling stage. SOCRAT/V3 has been used for estimation of further steps in experimental procedure to reach a representative LOCA scenario in future tests

La Naturaleza de la Experiencia Estética

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George Herbert Mead

2001-04-01

Full Text Available El texto que presentamos en este número de Athenea Digital es "La Naturaleza de la Experiencia Estética" escrito por G.H.Mead en 1926, traducido e introducido por Estaban Laso y Anna Vitores. Como los mismos traductores explican en la introducción que precede al texto de Mead: El artículo ?La Naturaleza de la Experiencia Estética? (1926 que presentamos en este mismo número ?y que como la inmensa mayoría de los escritos de Mead es muy breve? se inscribe dentro de la orientación hacia la crítica cultural de su obra tardía. En tanto que parte de la brecha entre la esfera instrumental y la de los valores que caracteriza a la actividad y al pensamiento humano en la sociedad moderna, ofrece una buena oportunidad para pensar acerca de la tensión meadiana entre creatividad y normatividad. Y quizás también, para ir más allá y reflexionar sobre la tensión entre cambio y reproducción social, y sobre las formas en que podríamos ofrecer un concepto de acción y de experiencia no individualistas que permitieran articular esta tensión.

Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

Energy Technology Data Exchange (ETDEWEB)

Bae, Youngmin; Kim, Young-In; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

2015-05-15

KAERI (Korea Atomic Energy Research Institute) is developing a passive safety injection system (PSIS) to supply cold borated water into a reactor coolant system (RCS) without any operator actions or AC power under the occurrence of postulated design basis accidents. The PSIS consists of four independent trains, each of which is furnished with a gravity drained core makeup tank (CMT) and a safety injection tank (SIT). The CMT is designed to provide makeup and boration functions to the RCS during the early stage of a loss of coolant accident (LOCA) and a non-LOCA. In this paper, we investigate numerically the fluid mixing characteristics in the upper annular space of SMART, especially when single-phase natural circulation is formed between the CMT and RCS following a non-LOCA such as a main steam line break. In this paper, the fluid mixing characteristics in the upper annular space of SMART are investigated numerically when single-phase natural circulation is formed between the RCS and CMT during the early stage of a non-LOCA.

Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

International Nuclear Information System (INIS)

Bae, Youngmin; Kim, Young-In; Kim, Keung Koo

2015-01-01

KAERI (Korea Atomic Energy Research Institute) is developing a passive safety injection system (PSIS) to supply cold borated water into a reactor coolant system (RCS) without any operator actions or AC power under the occurrence of postulated design basis accidents. The PSIS consists of four independent trains, each of which is furnished with a gravity drained core makeup tank (CMT) and a safety injection tank (SIT). The CMT is designed to provide makeup and boration functions to the RCS during the early stage of a loss of coolant accident (LOCA) and a non-LOCA. In this paper, we investigate numerically the fluid mixing characteristics in the upper annular space of SMART, especially when single-phase natural circulation is formed between the CMT and RCS following a non-LOCA such as a main steam line break. In this paper, the fluid mixing characteristics in the upper annular space of SMART are investigated numerically when single-phase natural circulation is formed between the RCS and CMT during the early stage of a non-LOCA

Calculation of the frequency of excedence in Full Spectrum LOCA by FSR; Calculo de la Frecuencia de excedencia en Full Spectrum LOCA mediante metodologia ISA

Energy Technology Data Exchange (ETDEWEB)

Gomez Magan, J. J.; Queral Salazar, C.; Sanchez Perea, M.

2012-07-01

In this application LOCA sequences was taken into account the uncertainty in the size of rupture and the operator action times as cooling and depressurization through steam generators. The simulations were performed using the tool SCAIS, dynamically coupled with MAAP code.

An application of RELAP5/MOD3 to the post-LOCA long term cooling performance evaluation

International Nuclear Information System (INIS)

Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

1998-01-01

A realistic long-term calculation to be used in the post-LOCA long term cooling (LTC) analysis is described in this study, which was required to resolve the post-LOCA LTC issues including the concern on boric acid precipitation in the reactor core. The analysis scope is defined according to the LTC plan of UCN Units 3/4 and the plant calculation model are developed suitable to the LTC procedure. The LTC sequences following the cold leg small break LOCAs of 0.02 ft2 to 0.5 ft2 are calculated by RELAP5/ MOD3.2.2. Based on the calculation results, the establishment of shutdown cooling system entry condition and the behavior of boron transport are evaluated. The effect of model simplification is also investigated

LWR fuel cladding deformation in a LOCA and its interaction with the emergency core cooling

International Nuclear Information System (INIS)

Erbacher, F.J.

1982-01-01

The paper summarizes research results of out-of-pile burst tests, in-pile bursts tests, out-of-pile flooding tests and modeling work on fuel behavior in a LOCA performed at KfK: The dominant phenomena of the cladding deformation and failure have been clarified by experiments and can be modeled by computer codes. The burst and flooding tests performed up to now suggest that the coolability of the core under LOCA conditions can be maintained. (orig.) [de

Review of RIA and LOCA criteria for WWER fuel

International Nuclear Information System (INIS)

Hozer, Z.

2006-01-01

The RIA and LOCA fuel safety criteria are under revision in the international community of fuel suppliers, authorities and research organizations. The main criteria will be reviewed in the paper for WWER fuel. Experimental data on the fuel failure behaviour under reactivity-initiated accident (RIA) conditions produced in the last decade in French and Japanese test reactors indicated low failure enthalpy for high burnup fuel compared to fresh fuel. However the high burnup was not the only phenomenon influencing the fuel failure. The oxide scale on the external surface of the fuel rod, hydrogen content of the Zr cladding and the local hydriding seemed also be responsible for the failure at low enthalpy. Furthermore differences have been found between Western design fuel and Russian type WWER fuel. The burnup dependence of fuel failure for WWER fuel was found much less, probably due to the low oxidation during normal operational conditions compared to other PWRs. The recently published Vitanza and KAERI correlations for RIA failure enthalpy have been applied to 23 WWER tests. Experimental data from Russian IGR and BIGR reactors have been used. The calculations have shown that both burnup and cladding oxidation effects must be considered, however the pulse width dependence of failure enthalpy has not been confirmed. During loss of coolant accidents (LOCA) the peak cladding temperature and local oxidation criteria have to be met. The oxidation criterion is under discussion today in many laboratories. The AEKI carried out several experimental series with Zr1%Nb cladding used in WWER reactors. The paper will describe the main results of the tests and present the limit for ductile-brittle transition derived from ring compression test. The behaviour of Zr1%Nb (E110) and Zircaloy-4 claddings under LOCA conditions will be compared as well. (author)

Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

Energy Technology Data Exchange (ETDEWEB)

Lu, S; Streit, R D; Chou, C K

1980-01-01

This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10{sup -12}). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

International Nuclear Information System (INIS)

Lu, S.; Streit, R.D.; Chou, C.K.

1980-01-01

This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10 -12 ). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

Bioactive compounds content of chimarrão infusions related to the moisture of yerba maté (Ilex Paraguariensis leaves

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Deborah H. M. Bastos

2006-05-01

Full Text Available The aim of this study was to evaluate the effects of the processing stages of yerba maté (Ilex paraguariensis on the moisture content of the leaves and the efficiency of the aqueous extraction of some bioactive substances. Samples of yerba maté were analyzed for caffeine, phenolic acids (caffeic acid, 5-caffeoilquinic acid and flavonoids (quercetin, kaempferol and myricetin by HPLC equipped with a diode array detector. Processing widely influenced the caffeine and 5-caffeoilquinic acid content of the aqueous extract (p A erva mate (Ilex paraguariensis é a matéria prima para três tipos de bebidas largamente consumidas na América do Sul. Substâncias bioativas presentes neste produto como a cafeína e os ácidos clorogênicos têm recebido especial atenção da comunidade científica. O objetivo deste trabalho é avaliar o efeito do processamento da erva mate no teor de umidade das folhas e a eficiência da extração aquosa de algumas substâncias bioativas. Amostras de erva mate coletadas no Paraná, Brasil foram objeto deste estudo. Cafeína, ácidos fenólicos (ácido cafeico e ácido 5-cafeoilquinico e flavonóides (quercitina, miricetina e caempferol foram analisados por HPLC equipado com detector de arranjo de diodos. Os teores de ácido 5-cafeoilquinico e cafeína do extrato aquoso variam em função da etapa do processamento (p0,9. O ácido cafeico foi determinado em 45% das infusões obtidas das folhas secas e quercitina, miricetina e caempferol não foram detectados nesses extratos.

Determinação do perfil de compostos voláteis e avaliação do sabor e aroma de bebidas produzidas a partir da erva-mate (Ilex paraguariensis Volatile compounds profile and flavor analysis of yerba mate (Ilex paraguariensis beverages

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Carla Carolina Batista Machado

2007-06-01

Full Text Available Volatile compounds from green and roasted yerba mate were analyzed by gas chromatography/mass spectrometry and the flavor profile from yerba mate beverages was determined by descriptive quantitative analyses. The main compounds tentatively identified in green mate were linalool, alpha-terpineol and trans-linalool oxide and in roasted mate were (E,Z-2,4-heptadienal isomers and 5-methylfurfural. Green mate infusion was qualified as having bitter taste and aroma as well as green grass aroma while roasted mate was defined as having a smooth, slightly burnt aroma. The relationship between the tentatively identified compounds and flavor must be determined by olfatometric analysis.

La concepción sociosemiótica de la cultura aplicada al estudio de la naturaleza

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MSc.Roxana Cruz-Doimeadios

2015-10-01

Full Text Available El presente artículo parte de las reflexiones de importantes investigadores que han investigado sobre la cultura, particularmente lo concerniente a los símbolos y los significados culturales, los que sirven de base para la elaboración del concepto construcción simbólica de la naturaleza que se aporta, se realiza un análisis de la condición de la naturaleza como un texto construido culturalmente, desde este enfoque se proyecta una novedosa manera de interpretar algunos de los fundamentos sobre los que sustentan la relación que se establece entre la naturaleza y la sociedad utilizando los fundamentos de la Teoría Cultural y la Semiótica fundamentalmente, sin desechar los aportes ofrecidos por otros sistemas teóricos para la comprensión del tema.

The blowdown, refill and reflood phase during a LOCA. Survey of the main physical phenomena

International Nuclear Information System (INIS)

Reocreux, M.

1980-05-01

In this paper, the main physical phenomena occuring during a LOCA are reviewed. They are presented in a chronological order. For each phenomena, a detailed physical description is given followed by the review of the general modelling problems. For some of these phenomena, modelling details are given for critical flow, for two-phase flow and heat transfer, for critical heat flux and post critical heat flux heat transfer, for reflood and rewet heat transfer and in the survey on LOCA computation codes

A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

International Nuclear Information System (INIS)

Popov, N.; Bosevski, T.

1986-01-01

The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

Fobia social: Naturaleza, evaluación y tratamiento (2015)

OpenAIRE

Bados López, Arturo

2015-01-01

Se abordan diversos aspectos de la fobia social: naturaleza, edad de comienzo y curso, frecuencia, problemas asociados, génesis y mantenimiento, métodos e instrumentos de evaluación, y eficacia y utilidad clínica del tratamiento psicológico y farmacológico. Además, se ofrecen guías para aplicar los tratamientos psicológicos más eficaces.

Analysis of LOCA/LOECC with a non-stop CATHENA simulation

International Nuclear Information System (INIS)

Sabourin, G.; Huynh, H.M.

1997-01-01

This paper documents a new approach which simulates without interruption the blowdown and the post-blowdown portions of a LOCA/LOECC. The blowdown portion is simulated first with the pressures, enthalpies, and void fractions of the headers as boundary conditions. The transient inlet header flowrates are written to a file. The blowdown portion is then simulated again with the inlet header flowrates as boundary conditions. At the end of the blowdown, the flowrates are gradually changed to obtain the desired constant gas flowrate of the post-blowdown portion. This new approach was applied with CATHENA MOD3.5a Rev. 0 for a 20% reactor inlet header break coincident with a total loss of emergency core cooling injection. In summary, this paper shows a successful new approach where the blowdown and the post-blowdown portions of a large LOCA coincident with a total loss of emergency core cooling are simulated continuously. (author)

BWR 2 % main recirculation line break LOCA tests RUNs 915 and 920 without HPCS in ROSA-III program

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

1987-03-01

This report presents the experimental results of BWR LOCA integral tests, RUNs 915 and 920, which are performed in the ROSA-III program simulating 2 % main recirculation line break LOCA tests with and without pressure control system operation. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, four types of ECCS's, and steam and feedwater systems. The report presents (1) the experimental results of 2 % small break LOCA phanomena in the ROSA-III system and (2) the effects of the pressure control system on the LOCA phenomena. The pressure control system contributed to (A) prevent bulk flashing in the early blowdown phase, (B) early closure of MSIV by L2 level trip, (C) early actuation of ADS by L1 level trip. However, the core thermal responses of the two tests were similar because of the similar mass inventory in PV after the ADS actuation in both tests. (author)

El arte de andar en la naturaleza

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María José Arbeláez Grundmann

2018-01-01

Full Text Available El arte de andar en la naturaleza gira alrededor de la cuestión ética y política de, ¿cómo los hombres nos relacionamos con la naturaleza cuando recorremos caminos? La respuesta se nutre de la filosofía ecofeminista espiritual de Vandana Shiva, la cual considera, junto con comunidades indígenas, a la tierra como un ser vivo, con el cual el hombre tiene una conexión espiritual y no es solamente garante de su existencia mediante la explotación ilimitada. Se amplía la reflexión en torno a los modos de relacionarse el hombre con la comunidad de los seres vivos mediante la propuesta de la ética de la tierra elaborada por Carlos Alfredo Vargas. Como práctica artística y proyecto de investigación, el arte de andar es una experiencia vivencial cotidiana consciente en el caminar diferentes recorridos, durante la cual se establecen relaciones de diversos órdenes, explorando los procesos creativos a partir de sensaciones corporales gracias a la apertura del caminante a las fuerzas intempestivas que atraviesan y modifican su cuerpo, tomado como referencia el “cuerpo sin órganos”, categoría desarrollada por Gilles Deleuze y Félix Guattari. El andar diferentes territorios permite la resignificación simbólica de los lugares, no necesitando expresarse en materialidades; por el contrario, esta práctica es tan respetuosa que solamente deja huellas móviles y evanescentes.

Influence of LOCA simulating conditions on the variation of electrical characteristics of insulating materials

Energy Technology Data Exchange (ETDEWEB)

Okada, Sohei; Yoshikawa, Masato; Ito, Masayuki; Kusama, Yasuo; Yagi, Toshiaki

1982-12-01

The authors have examined the variation of insulation resistance when the sheets of insulating materials and cables were exposed to various LOCA simulating environment. This report describes the summarized results obtained so far for ethylene propylene rubber (EPR) which is important as an insulating material of cables. The samples used were an EPR sheet of standard compound ratio, 2 kinds of EPR sheets of practical compound ratio, 6 types of PH cables (fire-retardant, EPR insulated, chlorosulphonated polyethylene sheathed cable) produced for trial as reactor use, and 6 kinds of EPR sheets of the same composition as the cable core. To discuss the difference of insulation resistance change, the logarithmic mean of the ratio of 1 min values to initial insulation resistance rho/rhosub(o) was used. PWR LOCA-simulating environment was used, while the thermal aging in the air at 121 deg C for 7 days and 50 Mrad irradiation in the air at room temperature were given as the predeterioration. The effect of LOCA-simulation period in the simultaneous method without air, in which steam and radiation were given in parallel, the difference in the experimental results of cables and sheets, the effect of air, the comparison of the simultaneous method with the sequential method in which LOCA-simulating steam was applied after the irradiation in the air and the reverse sequential method (dielectric property measurements) are described. Under the existence of air, the sequential method seems to be a good simulation condition for the simultaneous method, though many experiments are required further.

Influence of LOCA simulating conditions on the variation of electrical characteristics of insulating materials

International Nuclear Information System (INIS)

Okada, Sohei; Yoshikawa, Masato; Ito, Masayuki; Kusama, Yasuo; Yagi, Toshiaki

1982-01-01

The authors have examined the variation of insulation resistance when the sheets of insulating materials and cables were exposed to various LOCA simulating environment. This report describes the summarized results obtained so far for ethylene propylene rubber (EPR) which is important as an insulating material of cables. The samples used were an EPR sheet of standard compound ratio, 2 kinds of EPR sheets of practical compound ratio, 6 types of PH cables (fire-retardant, EPR insulated, chlorosulphonated polyethylene sheathed cable) produced for trial as reactor use, and 6 kinds of EPR sheets of the same composition as the cable core. To discuss the difference of insulation resistance change, the logarithmic mean of the ratio of 1 min values to initial insulation resistance rho/rhosub(o) was used. PWR LOCA-simulating environment was used, while the thermal aging in the air at 121 deg C for 7 days and 50 Mrad irradiation in the air at room temperature were given as the predeterioration. The effect of LOCA-simulation period in the simultaneous method without air, in which steam and radiation were given in parallel, the difference in the experimental results of cables and sheets, the effect of air, the comparison of the simultaneous method with the sequential method in which LOCA-simulating steam was applied after the irradiation in the air and the reverse sequential method (dielectric property measurements) are described. Under the existence of air, the sequential method seems to be a good simulation condition for the simultaneous method, though many experiments are required further. (Wakatsuki, Y.)

Fuel-rod response during the large-break LOCA Test LOC-6

International Nuclear Information System (INIS)

Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

1981-01-01

The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

Compendium of ECCS [Emergency Core Cooling Systems] research for realistic LOCA [loss-of-coolant accidents] analysis: Final report

International Nuclear Information System (INIS)

1988-12-01

In the United States, Emergency Core Cooling Systems (ECCS) are required for light water reactors (LWRs) to provide cooling of the reactor core in the event of a break or leak in the reactor piping or an inadvertent opening of a valve. These accidents are called loss-of-coolant accidents (LOCA), and they range from small leaks up to a postulated full break of the largest pipe in the reactor cooling system. Federal government regulations provide that LOCA analysis be performed to show that the ECCS will maintain fuel rod cladding temperatures, cladding oxidation, and hydrogen production within certain limits. The NRC and others have completed a large body of research which investigated fuel rod behavior and LOCA/ECCS performance. It is now possible to make a realistic estimate of the ECCS performance during a LOCA and to quantify the uncertainty of this calculation. The purpose of this report is to summarize this research and to serve as a general reference for the extensive research effort that has been performed. The report: (1) summarizes the understanding of LOCA phenomena in 1974; (2) reviews experimental and analytical programs developed to address the phenomena; (3) describes the best-estimate computer codes developed by the NRC; (4) discusses the salient technical aspects of the physical phenomena and our current understanding of them; (5) discusses probabilistic risk assessment results and perspectives, and (6) evaluates the impact of research results on the ECCS regulations. 736 refs., 412 figs., 66 tabs

A simplified time-dependent recovery model as applied to RCP seal LOCAs

International Nuclear Information System (INIS)

Kohut, P.; Bozoki, G.; Fitzpatrick, R.

1991-01-01

In Westinghouse-designed reactors, the reactor coolant pump (RCP) seals constantly require a modest amount of cooling. This cooling function depends on the service water (SW) system. Upon the loss of the cooling function due to the unavailability of the SW, component cooling water system or electrical power (station blackout), the RCP seals may degrade, resulting in a loss-of-coolant accident (LOCA). Recent studies indicate that the frequency of the loss of SW initiating events is higher than previously thought. This change significantly increases the core damage frequency contribution from RCP seal failure. The most critical/dominant element in the loss of SW events was found to be the SW-induced RCP seal failure. For these potential accident scenarios, there are large uncertainties regarding the actual frequency of RCP seal LOCA, the resulting leakage rate, and time-dependent behavior. The roles of various recovery options based on the time evolution of the seal LOCA have been identified and taken into account in recent NUREG-1150 probabilistic risk assessment PRA analyses. In this paper, a consistent time-dependent recovery model is described that takes into account the effects of various recovery actions based on explicit considerations given to a spectrum of time- and flow-rate dependencies. The model represents a simplified approach but is especially useful when extensive seal leak rate and core uncovery information is unavailable

Embrittlement of pre-hydrided Zircaloy-4 by steam oxidation under simulated LOCA transients

Energy Technology Data Exchange (ETDEWEB)

Desquines, J., E-mail: jean.desquines@irsn.fr; Drouan, D.; Guilbert, S.; Lacote, P.

2016-02-15

During a Loss Of Coolant Accident (LOCA), the mechanical behavior of high temperature steam oxidized fuel rods is an important issue. In this study, as-received and pre-hydrided axial tensile samples were steam oxidized in a vertical furnace and water quenched in order to simulate a LOCA transient. The samples were then subjected to a mechanical test to determine the failure conditions. Two different rupture modes were evidenced; the first one associated to linear elastic fracture mechanics and the second one is associated to sample failure without applied load. The oxidized cladding fracture toughness was determined relying on intensive metallographic analysis. The sample failure conditions were then back predicted confirming that the main rupture parameters are well captured.

LA NATURALEZA CONTRADICTORIA DEL ESTADO MEXICANO

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Carlos Báez Silva

2002-01-01

Full Text Available En el artículo se aborda el tema de la naturaleza del Estado mexicano, partiendo de la teoría de la contradicción del Estado capitalista adaptada al caso nacional. Se critican las posiciones tradicionales y se sostiene que la contradicción esencial del Estado en México consistió en su labor modernizadora mediante prácticas autoritarias, puesto que se institucionalizó la incertidumbre, entendido esto no como el desconocimiento del nombre del ganador de un proceso electoral preestablecido, sino como el uso y abuso de la discrecionalidad con que se gobernaba en México, la cual se convirtió en una verdadera institución, en el sentido económico del término.

Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

International Nuclear Information System (INIS)

Nelson, C.F.; Gauthier, G.; Carlin, F.

1996-10-01

Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40 degrees C or 70 degrees C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased

Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

Energy Technology Data Exchange (ETDEWEB)

Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yurko, Joseph P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swindlehurst, Gregg [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-04-01

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

A LOCA analysis for AHWR caused by ECCS header rupture

International Nuclear Information System (INIS)

Chatterjee, B.; Gawai, Amol; Gupta, S.K.; Kushwaha, H.S.

2000-01-01

Loss of coolant accident (LOCA) analyses for the proposed 750 MWth Advanced Heavy Water Reactor (AHWR), initiated by the rupture of 8 inch NB ECCS header has been carried out. This paper narrates the description of AHWR and associated ECCS, postulated scenario with which the analyses is carried out, results, discussion and conclusion

Studies in Phebus reactor of fuel behaviour upon LOCA conditions

International Nuclear Information System (INIS)

Manin, A.; Del Negro, R.; Reocreux, M.

1980-09-01

The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

SB LOCA analyses for Krsko Full Scope Simulator verification

International Nuclear Information System (INIS)

Prosek, A.; Parzer, I.; Mavko, B.

2000-01-01

Nuclear power plant simulators are intended to be used for training and maintaining competence to ensure safe, reliable operation of nuclear power plants throughout the world. The simulator shall be specified to a reference unit and its performance validation testing shall be provided. In this study a small-break loss-of-coolant accident (SB LOCA) response of Krsko nuclear power plant (NPP) was calculated for full scope simulator verification. The investigation included five cases with varying the break size in the cold leg of reactor coolant system. The plant specific and verified RELAP5/MOD2 model of Krsko nuclear power plant (NPP), developed in the past for 1882 MWt power, was adapted for 2000 MWt power (cycle 17) including the model for replacement steam generators. The results showed that the plant system response to breaks with small break area was slower compared to breaks with larger break area. The core heatup occurred in most of the cases analyzed. The acceptance criteria for emergency core cooling system were also met. The predicted results of the SB LOCA analysis for Krsko NPP suggest that they may be used for verification of the Krsko Full Scope Simulator performance. (author)

A new high temperature deformation model for zircaloy clad ballooning under hypothetical LOCA conditions

International Nuclear Information System (INIS)

Brzoska, B.; Cheliotis, G.; Kunick, A.; Senski, G.

1977-01-01

Assuming Zircaloy clad ballooning occurs predominantly by thermal activated secondary creep, generally a power law is applied to describe the creep rate analytically. According to Norton the creep rate is taken as a power function of the cladding hoop stress multiplied by a numerical constant which is determined by the cladding structural properties and a Boltzmann factor including the creep activation energy, the gas constant and the cladding temperature respectively. As is well known, the stress exponent is not a constant value in the total range of LOCA stresses, but increases steadily with stress. This difficulty is avoided by introducing into the Norton law a plastic flow-factor including a limiting stress, which was derived by G. Senski using plastic crack models from Dugdale and Irwin. For LOCA applications the limiting stress is identified with the burst stress, which is experimentally determined. A total number of about 280 directly heated KWU burst tests including two types of experiments: (i) controlled temperature transient tests, (ii) creep rupture tests, are used to fit the burst stress of KWU zircaloy tubes simulating the whole range of LOCA temperatur

Prediction of Golden Time for Recovering the Safety Injection System in Severe LOCA Circumstances

International Nuclear Information System (INIS)

Yoo, Kwae Hwan; Kim, Dong Young; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun

2015-01-01

In this study, the core uncovery and RV failure according to LOCA break sizes were analyzed by using the MAAP4 code when safety injection system (SIS) was not operating normally. We predicted the golden time of SIS recovery for accomplishing the reactor cold shutdown and preventing RV failure. MAAP4 code was used for severe accident analysis. The LOCA simulations were performed with break size in order to predict the golden time to recovery SIS. We predicted the golden time according to the SIS operation cases through the simulation of OPR1000. When LOCA occurred, the normal operation of SIS is very important in maintaining the integrity of NPPs. However if the SIS does not work or its actuation is delayed due to failure of the equipment, the DBA will lead to a severe accident. In this study, accident situations that SIS does not work normally were assumed and a number of MAAP4 code simulations were conducted. In addition, core uncovery time and RV failure time were predicted. If the recovery time of SIS for accident recovery is predicted, the core will not be exposed through appropriate action

ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA

International Nuclear Information System (INIS)

Araya, F.; Akimoto, M.

1985-01-01

1 - Nature of physical problem solved: ALARM-B2 which is an improved version of ALARM-B1 is a computer program to analyze thermo-hydraulic phenomena of BWR during a blowdown period under a large-break loss-of-coolant accident condition with special emphasis on the heat transfer phenomena in the core region. 2 - Method of solution: A so called volume-junction method is used to present fluid conservations. The primary system is divided into a number of special elements called 'control-volumes'. The system of partial differential equations describing fluid conservations for a stream-tube are integrated over a number of control volumes. The resulting set of simultaneous differential equations that is based on the assumptions of one-dimensional, homogeneous and thermal- equilibrium flow is linearized and solved for a small time increment by a simple explicit numerical technique. The one-dimensional heat conduction equations describing temperature profiles within solid material are written in finite difference forms which are linearized and solved by the Crank-Nicholson implicit method. In order to simulate the blowdown heat transfer phenomena, the code has correlation packages for heat transfer coefficient and critical heat flux. The heat generation in the core is given by a point reactor kinetics model with six groups of delayed neutrons and decay of eleven groups of fission products and actinides. The solution technique of the reactor kinetics is based on the Runge-Kutta method. ALARM-B2 has the models to simulate various components incorporated in BWRs such as jet pumps, recirculation pumps, steam separators, valves, and so on. The discharge and injection systems are modeled by leak and fill systems, respectively. 3 - Restrictions on the complexity of the problem: As this has been developed to simulate a blowdown thermo-hydraulic transient during a large break LOCA, users must pay attention when applying the code to any medium or small break LOCAs or to later phases

Potential for boron dilution during small-break LOCAs in PWRs

International Nuclear Information System (INIS)

Nourbakhsh, H.P.; Cheng, Z.

1995-01-01

This paper documents the results of a scoping study of boron dilution and mixing phenomena during small break loss of coolant accidents (LOCAs) in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler condenser mode. A problem may occur when the subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core were examined in this report. A bounding evaluation of the range of boron concentration entering the core during a small break LOCA in a typical Westinghouse-designed, four-loop plant is also presented in this report

Los significados culturales sobre la naturaleza en el asentamiento poblacional Siboney. Santiago de Cuba

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MSc. Roxana Cruz-Doimeadios

2015-10-01

Full Text Available Se realiza un estudio sobre los significados otorgados a la naturaleza después del evento meteorológico Sandy que azotó el país en el mes de octubre del año 2012 en el asentamiento poblacional Siboney, ubicado en el municipio Santiago de Cuba. Con anterioridad, los significados estaban asociados a percepciones de seguridad, tanto de nesecidades materiales como espirituales. Los resultados muestran la diversidad que caracteriza los significados culturales, los que se diferencian entre los actores individuales y los grupos, estas diferencias se reflejan en las acciones que desarrollan a través de las estrategias para el manejo de la naturaleza en el asentamiento.

La función del estado de naturaleza en El origen de la desigualdad entre los hombres

Directory of Open Access Journals (Sweden)

Prieto, Martín

2011-12-01

Full Text Available En este trabajo pretendo exponer y analizar la figura teórica del estado de naturaleza tal como es presentada en El origen de la desigualdad entre los hombres de Jean Jacques Rousseau, para determinar qué intención tiene y qué papel juega en el desarrollo del argumento de esta obra, y en su relación con el propósito general que allí comienza y que realizará sobre todo en El contrato social. Argumentaré la inviabilidad de interpretaciones de corte histórico, antropológico y psicológico, debilitando la importancia del aspecto descriptivo sobre la naturaleza humana que aun está presente en el texto, y concluiré que el principal impulso de esta figura teórica, construida filosóficamente, es el de proponer las bases argumentativas para un modelo político, y su contenido ha de tomarse como la disposición de los elementos prescriptivos que darán forma a su ambición ética y política. Más concretamente, argumentaré en contra de la posición de John Scott, quien sostiene una interpretación de tipo psicologista del Segundo Discurso, pero a la vez sugiero que la naturaleza prescriptiva de este tratado no anula el valor especulativo del mismo. Para ello, utilizaré la teoría de Aristóteles de la poesía y las consideraciones de Ricoeur sobre la narración para realizar una defensa de la capacidad de la figura del estado de naturaleza de ofrecer una imagen de la naturaleza humana en relación a sus posibilidades políticas, basada en otra forma de entender la historia y la filosofía política del iusnaturalismo.

Development of loca calculation capability with relap5-3D in accordance with the evaluation model methodology

International Nuclear Information System (INIS)

Liang, T.K.S.; Huan-Jen, Hung; Chin-Jang, Chang; Lance, Wang

2001-01-01

In light water reactors, particularly the pressurized water reactor (PWR), the severity of a LOCA (loss of coolant accident) will limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the PCT (peak cladding temperature) evaluation during LOCA, it generally takes more resources to develop. Instead, implementation of evaluation models required by the Appendix K of 10 CFR 50 upon an advanced thermal-hydraulic platform can also enlarge significant margin between the highest calculated PCT and the safety limit of 2200 F. The compliance of the current RELAP5-3D code with Appendix K of 10 CFR50 has been evaluated, and it was found that there are ten areas where code assessment and/or further modifications were required to satisfy the requirements set forth in the Appendix K of 10 CFR 50. The associated models for LOCA consequent phenomenon analysis should follow the major concern of regulation and be expected to give more conservative results than those by the best-estimate methodology. They were required to predict the decay power level, the blowdown hydraulics, the blowdown heat transfer, the flooding rate, and the flooding heat transfer. All of the ten areas included in above classified simulations have been further evaluated and the RELAP5-3D has been successfully modified to fulfill the associated requirements. In addition, to verify and assess the development of the Appendix K version of RELAP5-3D, nine separate-effect experiments were adopted. Through the assessments against separate-effect experiments, the success of the code modification in accordance with the Appendix K of 10 CFR 50 was demonstrated. We will apply another six sets of integral-effect experiments in the next step to assure the integral conservatism of the Appendix K version of RELAP5-3D on LOCA licensing evaluation. (authors)

Trastorno obsesivo compulsivo: naturaleza, evaluación y tratamiento (2015)

OpenAIRE

Bados López, Arturo

2015-01-01

Se abordan diversos aspectos del trastorno obsesivo-compulsivo: naturaleza, edad de comienzo y curso, frecuencia, problemas asociados, génesis y mantenimiento, métodos e instrumentos de evaluación, y eficacia y utilidad clínica del tratamiento psicológico y farmacológico. Además, se ofrecen guías para aplicar los tratamientos psicológicos más eficaces.

Trastorno obsesivo compulsivo: naturaleza, evaluación y tratamiento (2009)

OpenAIRE

Bados López, Arturo

2009-01-01

Se abordan diversos aspectos del trastorno obsesivo-compulsivo: naturaleza, edad de comienzo y curso, frecuencia, problemas asociados, génesis y mantenimiento, métodos e instrumentos de evaluación, y eficacia y utilidad clínica del tratamiento psicológico y farmacológico. Además, se ofrecen guías para aplicar los tratamientos psicológicos más eficaces.

LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressure injection System (HPIS)

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The sixth OECD LOFT experiment was conducted on 5 March 1984. It simulated a 1.8-in cold leg break LOCA with no HPIS available. This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncover and addressed accumulator injection at low pressure differentials. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

Los museos de historia natural en el siglo XIX: templos, laboratorios y teatros de la naturaleza

Directory of Open Access Journals (Sweden)

López-Ocón Cabrera, Leoncio

1999-08-01

Full Text Available Not available.Se explica en este texto cómo y por qué los museos de historia natural se transformaron en «catedrales de la ciencia» durante el siglo XIX. Se plantea que esos lugares de la ciencia interesaron y fascinaron a expertos y profanos porque durante ese tiempo histórico lograron conciliar una triple función cognoscitiva y comunicativa: la de ser templos, laboratorios y teatros de la naturaleza. 1 Como depósito de las colecciones del mundo natural se convirtieron en un lugar de acumulación de tesoros y de adhesión a valores políticos y científicos. 2 Como ámbito de trabajo de los naturalistas que revelaban las claves del orden natural se transformaron en centros de investigaciones para ordenar y descifrar la naturaleza. 3 Como escenarios en los que se mostraban las maravillas de la naturaleza se autoconcibieron como una especie de gran teatro en el que se representaban lecciones de cosas del mundo natural a modo de un libro abierto.

A coupled hydrodynamic-hydrochemical modeling for predicting mineral transport in a natural acid drainage system.

Science.gov (United States)

Zegers Risopatron, G., Sr.; Navarro, L.; Montserrat, S., Sr.; McPhee, J. P.; Niño, Y.

2017-12-01

The geochemistry of water and sediments, coupled with hydrodynamic transport in mountainous channels, is of particular interest in central Chilean Andes due to natural occurrence of acid waters. In this paper, we present a coupled transport and geochemical model to estimate and understand transport processes and fate of minerals at the Yerba Loca Basin, located near Santiago, Chile. In the upper zone, water presentes low pH ( 3) and high concentrations of iron, aluminum, copper, manganese and zinc. Acidity and minerals are the consequence of water-rock interactions in hydrothermal alteration zones, rich in sulphides and sulphates, covered by seasonal snow and glaciers. Downstream, as a consequence of neutral to alkaline lateral water contributions (pH >7) along the river, pH increases and concentration of solutes decreases. The mineral transport model has three components: (i) a hydrodynamic model, where we use HEC-RAS to solve 1D Saint-Venant equations, (ii) a sediment transport model to estimate erosion and sedimentation rates, which quantify minerals transference between water and riverbed and (iii) a solute transport model, based on the 1D OTIS model which takes into account the temporal delay in solutes transport that typically is observed in natural channels (transient storage). Hydrochemistry is solved using PHREEQC, a software for speciation and batch reaction. Our results show that correlation between mineral precipitation and dissolution according to pH values changes along the river. Based on pH measurements (and according to literature) we inferred that main minerals in the water system are brochantite, ferrihydrite, hydrobasaluminite and schwertmannite. Results show that our model can predict the transport and fate of minerals and metals in the Yerba Loca Basin. Mineral dissolution and precipitation process occur for limited ranges of pH values. When pH values are increased, iron minerals (schwertmannite) are the first to precipitate ( 2.5

Naturaleza jurídica del Ministerio Público

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Zamyr Vega Gutiérrez

2000-08-01

Full Text Available El Ministerio Público es un ente al servicio de la función judicial, pero autónomo y sujeto a la Constitución Política y a su Ley Orgánica. El Ministerio Público debe garantizar el "Debido Proceso Legal", entendido como el cumplimiento de los requisitos constitucionales en materia de procedimiento. Por ello, es vital realizar las reformas pertinentes a nuestras normas sustantivas y adjetivas en materia de Derecho penal y Procesal Penal, pues la creación del Ministerio Público tiene como objetivo descentralizar las funciones de acusar, defender y juzgar, que son propias del Sistema Inquisitivo que, por su naturaleza persecutoria, se opone al proceso de garantías.

Parameter study on the influence of prepressurization on LWR fuel rod behaviour during normal operation and hypothetical LOCA

International Nuclear Information System (INIS)

Fuchs, H.P.; Brzoska, B.; Depisch, F.; Sauermann, W.

1978-01-01

To analyse the influence of prepressurization on fuel rod behaviour, a parametric study has been performed considering the effects of the as-fabricated fuel rod internal prepressure on the normal operation and postulated LOCA red behaviour of a 1300 MWe1 KWU standard nuclear power plant pressurized water reactor. A reduction of prepressurization in the analysed range results in a negligible worsened normal operation behaviour whereas the LOCA behaviour is improved significantly. (author)

Yerba mate, coffee and tobacco: their possible neuroprotecting role in Uruguayan patients suffering from Parkinson Disease, Montevideo, 2016

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Lucía Amestoy

2017-11-01

Full Text Available Diversos reportes internacionales han demostrado que existen factores genéticos y ambientales que podrían predisponer al desarrollo de la Enfermedad de Parkinson (EP. Dichos estudios muestran la existencia de una relación inversa entre el consumo de tabaco y café y la EP, viéndose una relación dependiente de la dosis. La Yerba Mate (YM también ha mostrado, a nivel regional, dicha relación entre el riesgo de desarrollar EP y el consumo de YM. El objetivo es evaluar la posible asociación neuroprotectora entre el hábito de consumo prolongado y sostenido de la yerba mate y/o café y/o tabaco y el riesgo de desarrollar la EP. Para ello se realizó un estudio caso-control entre individuos con EP y controles sin EP. La exposición se midió el consumo e intensidad de YM, café y de tabaco. Se obtuvieron sus respectivos Odds Ratio (OR y su análisis en conjunto mediante regresión logística. Se analizaron los resultados de 128 casos y 256 controles. Tanto el consumo de YM como de tabaco, evidenció una relación inversa entre su consumo y el riesgo de desarrollar EP, OR: 0.56 (IC 95%, 0.33-0.95; p= 0.032 y OR: 0.46 (IC 95%, 0.25-0.85; p=0.013, respectivamente. El consumo de café se mostró como factor de riesgo para EP: OR: 2.04 (IC 95%, 1.21-3.44; p=0.007. El análisis multivariado muestra que la única variable que se mantiene como factor de riesgo fue el café. Concluimos que este trabajo, confirma en general buena parte de los datos ya reportados regional y mundialmente en cuanto a factores neuroprotectores ambientales de la EP.

Appetite Suppression and Antiobesity Effect of a Botanical Composition Composed of Morus alba, Yerba mate, and Magnolia officinalis.

Science.gov (United States)

Yimam, Mesfin; Jiao, Ping; Hong, Mei; Brownell, Lidia; Lee, Young-Chul; Hyun, Eu-Jin; Kim, Hyun-Jin; Kim, Tae-Woo; Nam, Jeong-Bum; Kim, Mi-Ran; Jia, Qi

2016-01-01

Background . Obesity and its comorbidities continue to challenge the world at an alarming rate. Although the long term solution lies on lifestyle changes in the form of dieting and exercising, drug, medical food, or dietary supplement interventions are required for those who are already obese. Here we describe a standardized blend composed of extracts from three medicinal plants: Morus alba , Yerba mate , and Magnolia officinalis for appetite suppression and metabolic disorders management. Method . Extracts were standardized to yield a composition designated as UP601. Appetite suppression activity was tested in acute feed intake rat model. Efficacy was evaluated in C57BL/6J mouse models treated with oral doses of 1.3 g/kg/day for 7 weeks. Orlistat at 40 mg/kg/day was used as a positive control. Body compositions of mice were assessed using a dual energy X-ray absorptiometry (DEXA). ELISA was done for insulin, leptin, and ghrelin level quantitation. Nonalcoholic steatohepatitis (NASH) scoring was conducted. Results . Marked acute hypophagia with 81.8, 75.3, 43.9, and 30.9% reductions in food intake at 2, 4, 6, and 24 hours were observed for UP601. Decreases in body weight gain (21.5% compared to the HFD at weeks 7 and 8.2% compared to baseline) and calorie intake (40.5% for the first week) were observed. 75.9% and 46.8% reductions in insulin and leptin, respectively, 4.2-fold increase in ghrelin level, and reductions of 18.6% in cholesterol and 59% in low-density lipoprotein were documented. A percentage body fat of 18.9%, 47.8%, 46.1%, and 30.4% was found for mice treated with normal control, HFD, Orlistat, and UP601, respectively. 59.3% less mesenteric fat pad and improved NASH scores were observed for UP601. Conclusion . UP601, a standardized botanical composition from Morus alba , Yerba mate , and Magnolia officinalis could be used as a natural alternative for appetite suppression, maintaining healthy body weight and metabolism management.

Appetite Suppression and Antiobesity Effect of a Botanical Composition Composed of Morus alba, Yerba mate, and Magnolia officinalis

Directory of Open Access Journals (Sweden)

Mesfin Yimam

2016-01-01

Full Text Available Background. Obesity and its comorbidities continue to challenge the world at an alarming rate. Although the long term solution lies on lifestyle changes in the form of dieting and exercising, drug, medical food, or dietary supplement interventions are required for those who are already obese. Here we describe a standardized blend composed of extracts from three medicinal plants: Morus alba, Yerba mate, and Magnolia officinalis for appetite suppression and metabolic disorders management. Method. Extracts were standardized to yield a composition designated as UP601. Appetite suppression activity was tested in acute feed intake rat model. Efficacy was evaluated in C57BL/6J mouse models treated with oral doses of 1.3 g/kg/day for 7 weeks. Orlistat at 40 mg/kg/day was used as a positive control. Body compositions of mice were assessed using a dual energy X-ray absorptiometry (DEXA. ELISA was done for insulin, leptin, and ghrelin level quantitation. Nonalcoholic steatohepatitis (NASH scoring was conducted. Results. Marked acute hypophagia with 81.8, 75.3, 43.9, and 30.9% reductions in food intake at 2, 4, 6, and 24 hours were observed for UP601. Decreases in body weight gain (21.5% compared to the HFD at weeks 7 and 8.2% compared to baseline and calorie intake (40.5% for the first week were observed. 75.9% and 46.8% reductions in insulin and leptin, respectively, 4.2-fold increase in ghrelin level, and reductions of 18.6% in cholesterol and 59% in low-density lipoprotein were documented. A percentage body fat of 18.9%, 47.8%, 46.1%, and 30.4% was found for mice treated with normal control, HFD, Orlistat, and UP601, respectively. 59.3% less mesenteric fat pad and improved NASH scores were observed for UP601. Conclusion. UP601, a standardized botanical composition from Morus alba, Yerba mate, and Magnolia officinalis could be used as a natural alternative for appetite suppression, maintaining healthy body weight and metabolism management.

El turismo de naturaleza en espacios naturales. El caso del parque regional de las Salinas y Arenales de San Pedro del Pinatar

OpenAIRE

Ballesteros Pelegrín, Gustavo Alfonso

2014-01-01

Los espacios naturales atraen a turistas que buscan el contacto con la naturaleza, surgiendo nuevos productos turísticos. Surgen, de este modo, nuevos productos turísticos. En la oferta de turismo de naturaleza del Parque Regional de las Salinas y Arenales de San Pedro del Pinatar destaca: deportes en la naturaleza, aventura y ecoturismo. Este último se divide en: turismo ornitológico y fotográfico, que junto al tradicional de sol y playa y de salud, se ve favorecido por su situación geográfi...

Effect of air on speed of insulating material deterioration under simulated LOCA environment. [Gamma radiation

Energy Technology Data Exchange (ETDEWEB)

Kusama, Yasuo; Yagi, Toshiaki; Ito, Masayuki; Okada, Sohei; Yoshikawa, Masato (Japan Atomic Energy Research Inst., Takasaki, Gunma. Takasaki Radiation Chemistry Research Establishment)

1982-12-01

To examine the quality approval testing method for the electric cables used for nuclear reactors, various covering insulating materials employed for the cables have been investigated from all angles. The factors which are considered to affect the deterioration of cable materials in a simulated LOCA (loss of coolant accident) environmental test are numerous. This paper reports on the result of investigation on the effect of air on the rate of deterioration of various organic materials usually used as the insulating and covering materials for the cables. Five kinds of polymer sheets (1 mm thick) used for reactor cables were employed as samples. The samples of both standard compounding ratio and the compounding ratio for practical reactor use were tested. As the deterioration prior to LOCA simulation, the thermal deterioration corresponding to 40 years aging (at 121 deg C for 7 days) was given, and subsequently, 50 Mrad gamma -irradiation at 1 Mrad/h was performed in the air. After that, the samples were subject to LOCA simulated environment. Since the results were different according to the kinds of samples, those are described separately for Hypalon, ethylene propylene rubber, cross-linked polyethylene, chloroprene and silicone rubber. The existence of air under LOCA environment accelerated the deterioration of insulation materials except silicone rubber, though its influence differed to the polymers. These materials swelled in the presence of air, and the degree of swelling increased with the temperature, having the close relation to oxidation deterioration. Polyethylene was more susceptible to the effect of air, and silicone rubber was rather stable. The samples of fire-retardant compounding ratio more swelled by water absorption than those of standard compounding ratio.

Assessment of CONTAIN and MELCOR for performing LOCA and LOVA analyses in ITER

International Nuclear Information System (INIS)

Merrill, B.J.; Hagrman, D.L.; Gaeta, M.J.; Petti, D.A.

1994-09-01

This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe accident thermal hydraulic analyses. This code will require the least modification to be appropriate for calculating thermal hydraulic behavior in ITER relevant conditions that include vacuum, cryogenics, ITER temperatures, and the presence of a liquid metal test module. The assessment of the aerosol transport models in these codes concludes that several modifications would have to be made to CONTAIN and/or MELCOR to make them applicable to the aerosol transport part of severe accident analysis in ITER

Yerba mate tea and mate saponins prevented azoxymethane-induced inflammation of rat colon through suppression of NF-kB p65ser(311) signaling via IkB-a and GSK-3ß reduced phosphorylation

Science.gov (United States)

Yerba mate tea (YMT) has a chemopreventive role in a variety of inflammatory diseases. The objective was to determine the capability of YMT and mate saponins to prevent azoxymethane (AOM)-induced colonic inflammation in rats. YMT (2% dry leaves, w/v, as a source of drinking fluid) (n = 15) and mat...

Transient analysis of blowdown thrust force under PWR LOCA

International Nuclear Information System (INIS)

Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

1982-10-01

The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

International Nuclear Information System (INIS)

Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

2004-01-01

As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

Memoria de los Andes, memoria de la naturaleza

OpenAIRE

Usselmann, Pierre

2013-01-01

Hace más de 35 años, terminando su tesis (1965), OlivierDollfus destacó la importancia de los Andes, “inmenso ysuntuoso campo”, para futuras investigaciones. En ese entonces sus estudios trataban de la geomorfología; escribía que “una de sus razones de ser es justamente este vaivén constante entre la observación de los procesos actuales y el intento de explicación, es decir el análisis de las formas heredadas” (1964). Esta preocupación, esta noción de memoria de la naturaleza, complementaria ...

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

Progress in realistic LOCA analysis

International Nuclear Information System (INIS)

Young, M.Y.; Bajorek, S.M.; Ohkawa, K.

2004-01-01

In 1988 the USNRC revised the ECCS rule contained in Appendix K and Section 50.46 of 10 CFR Part 50, which governs the analysis of the Loss Of Coolant Accident (LOCA). The revised regulation allows the use of realistic computer models to calculate the loss of coolant accident. In addition, the new regulation allows the use of high probability estimates of peak cladding temperature (PCT), rather than upper bound estimates. Prior to this modification, the regulations were a prescriptive set of rules which defined what assumptions must be made about the plant initial conditions and how various physical processes should be modeled. The resulting analyses were highly conservative in their prediction of the performance of the ECCS, and placed tight constraints on core power distributions, ECCS set points and functional requirements, and surveillance and testing. These restrictions, if relaxed, will allow for additional economy, flexibility, and in some cases, improved reliability and safety as well. For example, additional economy and operating flexibility can be achieved by implementing several available core and fuel rod designs to increase fuel discharge burnup and reduce neutron flux on the reactor vessel. The benefits of application of best estimate methods to LOCA analyses have typically been associated with reductions in fuel costs, resulting from optimized fuel designs, or increased revenue from power upratings. Fuel cost savings are relatively easy to quantify, and have been estimated at several millions of dollars per cycle for an individual plant. Best estimate methods are also likely to contribute significantly to reductions in O and M costs, although these reductions are more difficult to quantify. Examples of O and M cost reductions are: 1) Delaying equipment replacement. With best estimate methods, LOCA is no longer a factor in limiting power levels for plants with high tube plugging levels or degraded safety injection systems. If other requirements for

Special LOFT features for improved monitoring and survival of LOCA transients

International Nuclear Information System (INIS)

Goodrich, L.D.; Leach, L.P.; Klingler, T.B.; Morrow, J.C.; Phoenix, W.C.; Satterwhite, D.G.; Sumpter, K.C.; Rouhani, S.Z.; Welland, H.J.

1980-01-01

LOFT is designed to monitor and survive Loss-Of-Coolant-Accidents (LOCAs). This report presents the primary design difference from LPWRs that were required to accomplish this. These design differences may be of interest to the nuclear power generator industry. This report should be revised semi-annually or as developments in the LOFT Program require

INTERCONEXIÓN HOMBRE-MENTE-NATURALEZA DESDE EL TAOÍSMO UNA MIRADA DESDE EL YIN-YANG

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José Arlés Gómez Arévalo

2006-01-01

Full Text Available El presente artículo, corresponde al proyecto investigativo “Ciencia-espiritualidad”, que indaga, entre otros aspectos, por las conexiones que se establecen entre el mundo de la ciencia occidental y las antiguas tradiciones del Lejano Oriente, más concretamente por el aporte del hinduismo, del budismo y del taoísmo en la comprensión de las problemáticas del hombre contemporáneo. En esta investigación, se hace un análisis del taoísmo, tradición de la antigua civilización China y sus conexiones con la reflexión occidental sobre la energía, la mente humana y la naturaleza. En el pensamiento taoísta, hay una visión de unidad indisoluble entre hombre y naturaleza. El hombre mismo hace parte de la unidad del cosmos: él mismo es naturaleza. Desde esta perspectiva, se analizan algunos de los aportes que esta milenaria tradición hace a la investigación actual sobre ciencia y espiritualidad y a la comprensión de nuevos discursos que se constituyen en temas fronterizos entre la fe y la ciencia contemporánea.

In-pile experiments on fuel rod behavior during a LOCA

International Nuclear Information System (INIS)

Karb, E.; Pruessmann, M.; Sepold, L.

1980-05-01

This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

Educar con ética y valores ambientales para conservar la naturaleza

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Adriana de Castro Cuéllar

2009-01-01

Full Text Available Pese a los esfuerzos por conservar los recursos naturales no se ha podido frenar el deterioro ambiental, debido, posiblemente, a que no logramos tener conciencia y actitud de respeto hacia la naturaleza. La enseñanza de valores ambientales desde la infancia es una forma de generar cambios de visión y de apreciación de la naturaleza. En este trabajo se analiza el contenido del libro de texto de Ciencias Naturales de quinto grado de primaria y realizamos observaciones sobre la conducta de alumnos de este grado en dos escuelas de San Cristóbal de las Casas, Chiapas. Se les entrevistó y pidió que respondieran un cuestionario para identificar si los contenidos del libro de texto ofrecen una enseñanza de valores éticos-ambientales, y verificar si el niño expresa dichos valores. El contenido del programa oficial y sus ejecuciones aportan información valiosa sobre la problemática ambiental; sin embargo, es ineficiente para promover valores ambientales en los niños como el respeto a la flora y fauna, y el manejo responsable del agua y desechos.

Establishment of the operating procedure to prevent boron precipitation during Post-LOCA long term cooling for Korean Westinghouse 3-loop NPPs

International Nuclear Information System (INIS)

Choi, Han Rim; Kwon, Tae Soon; Ban, Chang Hwan; Jeong, Jae Hoon; Lee, Young Jin.

1996-11-01

During post-LOCA LTC the increase of the excess reactivity for the extended fuel cycle should require increasing the RWST boron concentration in order to ensure core subcritical state. To quantify the concentration increment, the calculation methods was developed for the post-LOCA RCS/Sump mixed mean boron concentration, which applied for Kori 3 and 4 and Ulchin 1 and 2 of the Westinghouse 3-loop nuclear power plants in Korean. From the calculation results, the minimum boric acid concentrations increased of the RWST and accumulator were determined consideration of the convenient operation for operator on reloading. Boric acid concentrations of the RWST and the accumulators for Westinghouse 3-loop type plants were increased to meet the post-LOCA shutdown requirement for the long life cycles from 12 months to 18 months. To maintain LTC capability following a LOCA, the switchover time is examined using boron code of prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results showed that hot leg recirculation switchover times were shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3 and 4 and 8 hours from 18 hours for Ulchin 1 and 2, respectively. The flow path in the mode J for Kori 3 and 4 was recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1 and 2. (author). 2 tabs., 12 figs., 13 refs

La educación ambiental y su importancia en la relación sustentable: Hombre-Naturaleza- Territorio

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Ronald Fernando Quintana Arias

2017-01-01

Full Text Available Objetivo. Promover la importancia de la educación ambiental en el fortalecimiento de la relación sustentable Hombre-Naturaleza-Territorio en 8 niños y niñas observadores de aves de la primaria María Berchmans. Metodología. Entre agosto y noviembre del 2015 desarrollé un estudio de caso basado en el aprendizaje vivencial, social y experiencial, que se dividió en dos fases en las que se encontraban 19 actividades clasificadas dentro de siete categorías. Resultados. El desarrollo tanto de habilidades cognitivas “duras” como de habilidades socioemocionales “blandas” exponiendo diferentes ambientes educacionales, incentiva un aprendizaje activo que fortalece la relación sustentable Hombre-Naturaleza-Territorio. Conclusión. La generación de una conciencia de conservación ambiental al contemplar la importancia de la educación escolar en la relación Hombre-Naturaleza-Territorio lleva a la apropiación de la biodiversidad (natural-cultural y a generar aprendizajes significativos.

A study on the effect of the CHF correlations to the LOCA analysis

International Nuclear Information System (INIS)

Kim, Ho Kee

1998-02-01

The critical heat flux (CHF) is a major parameter which determines the cooling performance and therefore the prediction of CHF is of importance for the design and safety analysis in boiling systems; such as nuclear reactors, conventional boilers, and other various two-phase flow systems. Until now, many CHF correlations have been developed and for the actual design a correlation has been selected in consideration of its characteristics. For the analysis of Loss of Coolant Accident (LOCA) in a Nuclear Power Plant, which shows the drastic parameters change during the system transient, a correlation having a reasonable degree of accuracy over a wide range is preferred, rather than that having accuracy for a specific range. It is required to have tangible insight about the effects of the CHF correlation to the LOCA analysis for the purpose of computer code development and nuclear regulation. The related research is further recommended. The purpose of this research is to obtain an insight and/or intuition about the above effect and to evaluate the selected CHF correlations. To achieve these purposes LOCA is analysed for the UL-JIN 3 and 4 nuclear power plant, the Korea Standard Type Nuclear Power Plant and the Loss of Flow Test (LOFT) L2-5 experiment is simulated using the RELAP5/MOD3.1 computer code for each selected CHF correlation. The selected correlations are the AECL-UO Lookup Table, adapted in RELAP5 code; the K110 CHF correlation, developed by KAERI; and the original W-3 CHF correlation, developed by L.S. Tong. LOFT is also simulated using the AECL-UO Lookup Table having the CHF multiplication factors 0.5 and 1.5, and then compared with the result of the original Lookup Table and the experiment result. In the LOCA analysis, the CHF correlations affect the magnitude of peak cladding temperatures, but does not seriously affect the occurrence points of time. The effect of each CHF correlation to the fuel cladding temperature behavior becomes apparent at the end of

HISTORIA Y NATURALEZA DISCIPLINAR DE LA ASTROBIOLOGÍA

OpenAIRE

Octavio Alfonso Chon

2015-01-01

La astrobiología es una ciencia que se ha ido perfilándose dentro de la variedad de ciencias a partir de mediados de los años 90’. La astrobiología es el estudio, entre otros temas, del pasado, presente y futuro de la vida en la Tierra y en el universo. En el presente artículo se abordan el surgimiento de la definición de la palabra astrobiología, algunas consideraciones sobre la definición de vida y la naturaleza disciplinar de la astrobiología. El propósito de este trabajo es co...

Small break LOCA analysis for RCP trip strategy for YGN 3 and 4 emergency procedure guidelines

International Nuclear Information System (INIS)

Suh, Jong Tae; Bae, Kyoo Hwan

1995-01-01

A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3 and 4. The trip setpoint of the first two RCPs for YGN 3 and 4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 3 and 4 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at worst time. Also, the YGN 3 and 4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3 and 4 can provide improved operator guidance for the RCP operation during accidents. 11 figs., 4 tabs., 9 refs. (Author)

Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

International Nuclear Information System (INIS)

Griggs, D.P.

1993-03-01

The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

A through calculation of 1,100 MWe PWR large break LOCA by THYDE-P1 EM model

International Nuclear Information System (INIS)

Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

1984-07-01

THYDE-P1 is a code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs). Up to now, THYDE-P1 has been applied to various experiment analyses, which show its high capability to analyze LOCAs as a best estimate (BE) calculation code. In this report, evaluation model (EM) calculation method, especialy in the blowdown and refill phases, is established equivalently to WREM/J2 which is regarded as appropriate for an EM calculation code, and the results of them are compared and discussed. The present calculation was the first executed by THYDE-P1-EM, and was performed as Sample Calculation Run 80 which was a part of a series of THYDE-P sample calculations. The calculation was carried out from the LOCA initiation till 400 seconds for a guillotine break at the cold leg of a commercial 1,100 MWe PWR plant. The calculated results agreed well to that of the WREM/J2 code. (author)

Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code; Analisis de un accidente LOCA en contencion de un reactor PWR-W con el codigo GOTHIC

Energy Technology Data Exchange (ETDEWEB)

Perianez Alvarez, V.

2013-07-01

The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

Efecto nodriza intra-específico de Kageneckia angustifolia D. Don (Rosaceae sobre la germinación de semillas y sobrevivencia de plántulas en el bosque esclerófilo montano de Chile central Intra-specific nurse effect of Kageneckia angustifolia D. Don (Rosaceae and its effect on seed germination and seedling survival in the montane sclerophyllous forest of central Chile

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ALEJANDRO PEÑALOZA

2001-09-01

Full Text Available El bosque esclerófilo montano de Chile central (32-33° S, 1.500-2.100 m de altitud está dominado por poblaciones de Kageneckia angustifolia (Rosaceae, especie semidecidua de verano que forma un dosel muy abierto. Esto sugiere que, a diferencia de lo que ocurre en el matorral esclerófilo de menor altitud donde el cerrado dosel de árboles y arbustos generan condiciones microclimáticas diferentes a los espacios abiertos, en el bosque montano no existiría una marcada diferencia microclimática entre bajo el dosel y los espacios abiertos. Por otro lado, en el bosque montano, las precipitaciones ocurren principalmente en forma de nieve, la que se acumula preferentemente en los espacios entre los árboles, pudiendo facilitar el reclutamiento de nuevos individuos en este microhábitat, fenómeno que se conoce como efecto nodriza. Se estudió el probable efecto nodriza a nivel intra-específico de K. angustifolia comparando el microclima de los ambientes bajo dosel y los espacios abiertos, y el efecto de la acumulación de nieve en la germinación de semillas y sobrevivencia de plántulas de en un bosque esclerófilo montano ubicado en el Santuario de la Naturaleza Yerba Loca, 50 km al este de Santiago (33° S, 1.600 m de altitud. De acuerdo a las variables microclimáticas estudiadas (PAR, humedad del aire y suelo, y temperatura del aire y suelo, en el bosque montano no existen diferencias microclimáticas entre los espacios abiertos y bajo el dosel. Sólo la acumulación de nieve fue significativamente mayor en los espacios abiertos. La germinación fue menor y más tardía en los espacios abiertos, lo que estaría relacionado con la mayor acumulación de nieve. Las plántulas originadas más tempranamente tienen más tiempo para desarrollarse y pasar en forma exitosa la sequía estival en comparación con las plántulas que emergen más tardíamente. Esto explicaría la menor sobrevivencia de las plántulas en los espacios abiertosThe montane

Fission Product Transport Models Adopted in REFPAC Code for LOCA Conditions in PWR and WWER NPPS

International Nuclear Information System (INIS)

Strupczewski, A.

2003-01-01

The report presents assumptions and physical models used for calculations of fission product releases from nuclear reactors, their behavior inside the containment and leakages to the environment after large break loss of coolant accident LB LOCA. They are the basis of code REFPAC (RElease of Fission Products under Accident Conditions), designed primarily to represent significant physical processes occurring after LB LOCA. The code describes these processes using three different models. Model 1 corresponds to established US and Russian practice, Model 2 includes all conservative assumptions that are in agreement with the actual state-of-the-art, and Model 3 incorporates formulae and parameter values actually used in EU practice. (author)

R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

Energy Technology Data Exchange (ETDEWEB)

Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron Simon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tu, Lei [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-04-01

The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize

Modeling heat and mass transfer in the heat treatment step of yerba maté processing

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J. M. Peralta

2007-03-01

Full Text Available The aim of this research was to estimate the leaf and twig temperature and moisture content of yerba maté branches (Ilex paraguariensis Saint Hilaire during heat treatment, carried out in a rotary kiln dryer. These variables had to be estimated (modeling the heat and mass transfer due to the difficulty of experimental measurement in the dryer. For modeling, the equipment was divided into two zones: the flame or heat treatment zone and the drying zone. The model developed fit well with the experimental data when water loss took place only in leaves. In the first zone, leaf temperature increased until it reached 135°C and then it slowly decreased to 88°C at the exit, despite the gas temperature, which varied in this zone from 460°C to 120°C. Twig temperature increased in the two zones from its inlet temperature (25°C up to 75°C. A model error of about 3% was estimated based on theoretical and experimental data on leaf moisture content.

Sensitivity analysis of local uncertainties in large break loss-of-coolant accident (LB-LOCA) thermo-mechanical simulations

Energy Technology Data Exchange (ETDEWEB)

Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Ikonen, Timo

2016-08-15

Highlights: • A sensitivity analysis using the data from EPR LB-LOCA simulations is done. • A procedure to analyze such complex data is outlined. • Both visual and quantitative methods are used. • Input factors related to core design are identified as most significant. - Abstract: In this paper, a sensitivity analysis for the data originating from a large break loss-of-coolant accident (LB-LOCA) analysis of an EPR-type nuclear power plant is presented. In the preceding LOCA analysis, the number of failing fuel rods in the accident was established (Arkoma et al., 2015). However, the underlying causes for rod failures were not addressed. It is essential to bring out which input parameters and boundary conditions have significance to the outcome of the analysis, i.e. the ballooning and burst of the rods. Due to complexity of the existing data, the first part of the analysis consists of defining the relevant input parameters for the sensitivity analysis. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The ultimate goal is to develop a systematic procedure for the sensitivity analysis of statistical LOCA simulation that takes into account the various sources of uncertainties in the calculation chain. In the current analysis, the most relevant parameters with respect to the cladding integrity are the decay heat power during the transient, the thermal hydraulic conditions in the rod’s location in reactor, and the steady-state irradiation history of the rod. Meanwhile, the tolerances in fuel manufacturing parameters were found to have negligible effect on cladding deformation.

AEEW comments on the NNC/CEGB LOCA code validation report RX 440-A

International Nuclear Information System (INIS)

Brittain, I.; Bryce, W.M.; O'Mahoney, R.; Richards, C.G.; Gibson, I.H.; Porter, W.H.L.; Fell, J.

1984-03-01

Comments are made on the NNC/CEGB report PWR/RX 440-A, Review of Validation for the ECCS Evaluation Model Codes, by K.T. Routledge et al, 1982. This set out to review methods and models used in the LOCA safety case for Sizewell B. These methods are embodied in the Evaluation Model Computer codes SATAN-VI, WREFLOOD, WFLASH, LOCTA-IV and COCO. The main application of these codes is the determination of peak clad temperature and overall containment pressure. The comments represent the views of a group which has been involved for a number of years in the development and application of Best-Estimate methods for LOCA analysis. It is the judgement of this group that, overall, the EM methods can be used to make an acceptable safety case, but there are a number of points of detail still to be resolved. (U.K.)

Colonialidad de la naturaleza: ¿imposición tecnológica y usurpación epistémica? Interculturalidad, desarrollo y re-existencia

Directory of Open Access Journals (Sweden)

Adolfo Albán A.

2016-01-01

Full Text Available El artículo cuestiona las concepciones eurocéntricas de naturaleza que el proyecto moderno/ colonial instauró como racionalidad económica. Igualmente indaga sobre las formas de rela- cionamiento entre las sociedades culturalmente diferenciadas y la naturaleza biológicamente diversa. En consecuencia, plantea la necesidad de no naturalizar el relacionamiento cultural, sino entender la interculturalidad como un proyecto político, ético y epistémico, donde la edu- cación juega un papel fundamental para la transformación de paradigmas e imaginarios en torno al desarrollo y la naturaleza.

Effective water cooling of very hot surfaces during the LOCA accident.

Czech Academy of Sciences Publication Activity Database

Štepánek, J.; Bláha, V.; Dostál, V.; Entler, Slavomír

2017-01-01

Roč. 124, November (2017), s. 1211-1214 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : LOCA * Quenching * Divertor cooling * Heat transfer * Rewetting Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617303733

El turismo de naturaleza en espacios naturales. El caso del Parque Regional de las Salinas y Arenales de San Pedro del Pinatar

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Gustavo A. Ballesteros Pelegrín

2014-01-01

Full Text Available Los espacios naturales atraen a turistas que buscan el contacto con la naturaleza, surgiendo nuevos productos turísticos. Surgen, de este modo, nuevos productos turísticos. En la oferta de turismo de naturaleza del Parque Regional de las Salinas y Arenales de San Pedro del Pinatar destaca: deportes en la naturaleza, aventura y ecoturismo. Este último se divide en: turismo ornitológico y fotográfico, que junto al tradicional de sol y playa y de salud, se ve favorecido por su situación geográfica, condiciones naturales y amplia oferta hotelera. Sin embargo, el turismo cultural basado en la tradición salinera y pesquera no ha sido desarrollado.

Water volume available for ECCS sump recirculation mode following a LOCA

International Nuclear Information System (INIS)

Riekert, T.; Rebohm, H.; Huber, J.; Brandes, F.

2006-01-01

In this paper we describe the reviews performed in Germany on the water level in the containment sump after a LOCA and the derived actions. Our view on the issue is from the perspective of the independent safety experts - i.e. TUV SUD Industrie Service (TUV SUD IS), TUV SUD Energietechnik GmbH Baden-Wuerttemberg (TUV SUD ET), TUV NORD EnSys Hannover and TUV NORD SysTec -, which reviewed the analyses of the utilities on behalf of the responsible supervising authorities. Between these expert organizations information were exchanged via the steering committee on nuclear technology of the association of the TUVs (VdTUV). In our paper we describe the analyses on the two safety issues relevant in the connection with the water level in the containment sump: the necessary minimum coverage of suction pipes to avoid inadmissible entrainment of air and the water retention inside the containment after a LOCA. Our description concentrates on PWRs because of the more complex conditions in comparison to BWRs. In conclusion it can be stated that due to the thorough evaluation of operating experience, optimization measures could be derived. In addition, the analyses served the purpose of know-how maintenance. (authors)

Water volume available for ECCS sump recirculation mode following a LOCA

Energy Technology Data Exchange (ETDEWEB)

Riekert, T. [TUV NORD SysTec (Germany); Rebohm, H. [TUV NORD EnSys Hannover (Germany); Huber, J. [TUV SUD IS (Germany); Brandes, F. [TUV SUD ET (Germany)

2006-07-01

In this paper we describe the reviews performed in Germany on the water level in the containment sump after a LOCA and the derived actions. Our view on the issue is from the perspective of the independent safety experts - i.e. TUV SUD Industrie Service (TUV SUD IS), TUV SUD Energietechnik GmbH Baden-Wuerttemberg (TUV SUD ET), TUV NORD EnSys Hannover and TUV NORD SysTec -, which reviewed the analyses of the utilities on behalf of the responsible supervising authorities. Between these expert organizations information were exchanged via the steering committee on nuclear technology of the association of the TUVs (VdTUV). In our paper we describe the analyses on the two safety issues relevant in the connection with the water level in the containment sump: the necessary minimum coverage of suction pipes to avoid inadmissible entrainment of air and the water retention inside the containment after a LOCA. Our description concentrates on PWRs because of the more complex conditions in comparison to BWRs. In conclusion it can be stated that due to the thorough evaluation of operating experience, optimization measures could be derived. In addition, the analyses served the purpose of know-how maintenance. (authors)

Yerba Maté (Ilex paraguariensis) Metabolic, Satiety, and Mood State Effects at Rest and during Prolonged Exercise.

Science.gov (United States)

Alkhatib, Ahmad; Atcheson, Roisin

2017-08-15

Yerba Maté (YM), has become a popular herb ingested for enhancing metabolic health and weight-loss outcomes. No studies have tested the combined metabolic, satiety, and psychomotor effects of YM during exercise. We tested whether YM ingestion affects fatty acid oxidation (FAO), profile of mood state score (POMS), and subjective appetite scale (VAS), during prolonged moderate exercise. Twelve healthy active females were randomized to ingest either 2 g of YM or placebo (PLC) in a repeated-measures design. Participants rested for 120 min before performing a 30-min cycling exercise corresponding to individuals' crossover point intensity (COP). FAO, determined using indirect calorimetry, was significantly higher during the 30-min exercise in YM vs. PLC (0.21 ± 0.07 vs. 0.17 ± 0.06 g/min, p exercise at targeted "fat-loss"' intensities augments FAO and improves measures of satiety and mood state. Such positive combined metabolic, satiety, and psychomotor effects may provide an important role for designing future fat and weight-loss lifestyle interventions.

On-line pressurizer surveillance system design to prevent small break LOCA through PORV using micro-computer

International Nuclear Information System (INIS)

Lee, Jong-Ho; Chang, Soon-Heung

1986-01-01

Small break LOCA caused by a stuck-open PORV is one of the important contributors to nuclear power plant risk. This paper deals with the design of a pressurizer surveillance system using micro-computer to prevent the malfunction of system and has assessed the effect of this improvement through Probabilistic Risk Assessment (PRA) method. Micro-computer diagnoses the malfunction of system by a process checking method and performs automatically backup action related to each malfunction. Owing to this improvement, we can correctly diagnose ''Spurious Opening'', ''Fail to Reclose'' and ''Small break LOCA'' which are difficult for operator to diagnose quickly and correctly and reduce the probability of a human error by an automatic backup action. (author)

LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The third OECD LOFT experiment was conducted on 14 July 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with delayed pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The second OECD LOFT experiment was conducted on 23 June 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with early pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

Analysis of the bubble condenser structure of WWER-440 NPP under LOCA loading

International Nuclear Information System (INIS)

Zeman, P.

2003-01-01

Two problems may arise in relation to the title topic: (1) problem with the uplift of the beams I 600 of the first floor, and (2) possible plastic collapse of the wall on the 12th floor. The problems were tacked by computer calculations. The FEM model of the bubble condenser was created in the ANSYS 6.0 environment and analyzed for the pressure loading defined for the LOCA accident in IAEA TECDOC 803. The model of the bubble condenser structure so created included all geometrical and material non-linearities. The duration of the pressure wave was 0.4 s, amplitude 30 kPa. The analyses revealed that a plastic collapse of the tank wall is not the most critical failure mode. Instead, weld connections appear to be the most critical parts of structure. The tank walls are very ductile and the results of the analyses are in agreement with the test simulating the LOCA accident. The tank walls suffered no damage during the tests

Mixing phenomena of interest to boron dilution during small break LOCAs in PWRs

International Nuclear Information System (INIS)

Nourbakhsh, H.P.; Cheng, Z.

1995-01-01

This paper presents the results of a study of mixing phenomena related to boron dilution during small break loss of coolant accidents (LOCAs)in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler-condenser mode. A problem may occur when subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core are examined in this paper. Bounding calculations for boron concentration of coolant entering the core during a small break LOCA in a typical Westinghouse-designed four-loop plant are also presented

Radial heat transfer from fuel to moderator during LOCAs for CANDU PHW reactors

International Nuclear Information System (INIS)

Hildebrandt, J.G.; So, C.B.; Gillespie, G.E.; MacLean, G.

1983-01-01

In a postulated CANDU-PHW loss-of-coolant accident (LOCA) with coincident impaired emergency cooling, the axial transport of heat from the fuel by convection is reduced. This reduction in heat removal causes the fuel to heat up and the radial heat transfer to the moderator to become significant. This paper deals with two codes that predict the thermal response of fuel channels under LOCA conditions. New channel thermal radiation models in both RAMA, a thermalhydraulic code, and CHAN II, a fuel channel thermo-chemical code, are presented and their predictions are compared with the experimental results of an electrically heated bundle of 37 fuel pins. A second experiment, involving a single heated pin in a channel with flowing steam, is presented. The predictions of RAMA and CHAN II are compared with this experiment to verify the codes' thermo-chemical models. There is good agreement between the predictions of both codes and the experimental results

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

International Nuclear Information System (INIS)

Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

1995-10-01

This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken

Calculation of fuel pin failure timing under LOCA conditions

International Nuclear Information System (INIS)

Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

1991-10-01

The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

Large-break LOCA assessment for the highly advanced core design

International Nuclear Information System (INIS)

Doria, F.J.; Nath, V.I.; Hau, K.F.; Dam, R.F.; Vecchiarelli, J.

1997-01-01

Over the course of the years, a conceptual highly advanced core (HAC) reactor has been designed for Japan Electric Power Development Company Limited (EPDC). The HAC reactor, which is capable of generating 1326 MW of electrical power, consists of 640 CANDU-type fuel channels with each fuel channel containing twelve 61-element fuel bundles. As part of the conceptual design study, the performance of the HAC reactor during a large loss-of-coolant accident (LOCA) was assessed with the use of several computer codes. The SOPHT, CATHENA, ELOCA and ELESTRES computer codes were used to predict the thermalhydraulic behaviour of the circuit, thermalhydraulic behaviour of a single high-power channel, thermal-mechanical behaviour of the outer fuel elements contained in the high-powered channel and the steady-state fuel-element conditions respectively. The LOCAs that were analyzed include 100% reactor outlet header (ROH) break, and a survey of reactor inlet header (RIH) breaks ranging from 5% to 25%. The conceptual feasibility of the HAC design was evaluated against two criteria; namely, maximum sheath temperature less than 1200 deg C and AECL's 5% sheath straining criterion to assess failure by excessive straining. For the cases analyzed, the analysis predicted a maximum sheath temperature of 820 deg C and a maximum sheath strain of 1.5% (the maximum pressure-tube temperature was 515 deg C). Although the maximum element-burnup of the HAC design is extended beyond the CANDU 6 burnup, the maximum linear power of HAC (40 kW/m) is significantly lower than the maximum linear power of a CANDU 6 reactor (60 kW/m). The reduced element-power level in conjunction with internal design modification for the HAC design has resulted m significantly lower internal gas pressures under steady-state conditions, as compared with the CANDU 6 design. During a LOCA, the low linear powers and zero-void reactivity associated with the HAC design has increased the safety margin. In addition, the cases

High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

International Nuclear Information System (INIS)

Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

2013-01-01

In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

COLONIZACIÓN DE LA NATURALEZA: UNA APROXIMACIÓN DESDE EL EXTRACTIVISMO EN COLOMBIA.

Directory of Open Access Journals (Sweden)

Luis Alfredo Bohórquez Caldera

2013-06-01

Full Text Available This research paper aims at providing some elements of analysis on the complex issue of the colonization of nature. In it a descriptive argument is carried out based on theoretical findings on the relationships between extractivism, which is a proper practice of the colonial device, and what is here called colonization of nature. The article initially presents an assessment of the impact on the configuration of the ancestral vision and appropriation of the territory and its cultural contents, as a result of the imposition of a new semantics, in reference to the device of the colonial powers. Then, the relationship between extractivism and the process of colonization of nature is analyzed; to finally sketch a brief reflection, by linking contemporary issues and contexts. Este artículo de investigación pretende brindar elementos de análisis sobre el complejo tema de la colonización de la naturaleza. En él se hace una argumentación descriptiva basada hallazgos teóricos sobre las relaciones entre el extractivismo, una práctica propia del dispositivo colonial, y lo que aquí denominamos colonización de la naturaleza. El artículo presenta inicialmente un balance del impacto en la configuración de la visión-apropiación ancestral del territorio y su contenido cultural, a raíz de la imposición de una nueva semántica, en referencia al dispositivo de poder colonial. Luego, se analiza la relación del extractivismo con el proceso de colonización de la naturaleza, para finalmente esbozar una reflexión breve vinculando contextos y problemas contemporáneos.

LA NATURALEZA PROVEEDORA DE BIENESTAR Y BELLEZA, UN DISCURSO DESDE LA PERSPECTIVA LLANERA QUE ENRIQUECE LA CLASE DE CIENCIAS.

Directory of Open Access Journals (Sweden)

Andrés Arturo Venegas Segura

2014-05-01

Full Text Available Se presenta el discurso de Flor sobre la naturaleza, el cual resalta valores éticos, estéticos, emocionales, regulatorios enmarcados por un Ethos Llanero, donde el mundo natural es concebido como bello y ofrece utilidad, por tanto se debe respetar y conservar. Se utilizan los conglomerados de relevancias como dispositivo interpretativo de una carta al extraterrestre, un dibujo y una entrevista semi-estructurada, instrumentos que permitieron la recolección de la información. Palabras Clave: Idea de Naturaleza, conglomerados de relevancia, clase de ciencias naturales, Ethos Llaneros.

Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

Directory of Open Access Journals (Sweden)

F. Reventós

2012-01-01

Full Text Available Integral test facilities (ITFs are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results.

Fluid-structure coupled dynamic response of PWR core barrel during LOCA

International Nuclear Information System (INIS)

Lu, M.W.; Zhang, Y.G.; Shi, F.

1991-01-01

This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

Seminario sobre el proceso científico : naturaleza de la ciencia, investigación científica y relaciones ciencia-tecnología-sociedad

OpenAIRE

Bennàssar-Roig, A.; Vich-Sbert, M.A.; Manassero-Mas, M.A.; Vázquez-Alonso, A.; Tur-Marí, J.A

2017-01-01

Se presentan los resultados obtenidos en los cambios sobre concepciones científicas al realizar un seminario con estudiantes de grado de biología sobre el proceso científico: naturaleza de la ciencia, investigación científica y relaciones ciencia-tecnología-sociedad, de carácter teórico y práctico. Se realizó una sesión teórica sobre naturaleza de la ciencia y dos prácticas: lectura de un artículo para aplicar los contenidos de la naturaleza de la ciencia y análisis de un proyecto realizado q...

Seminario sobre el proceso científico : naturaleza de la ciencia, investigación científica y relaciones ciencia-tecnología- sociedad

OpenAIRE

Bennàssar-Roig, A.

2017-01-01

Se presentan los resultados obtenidos en los cambios sobre concepciones científicas al realizar un seminario con estudiantes de grado de biología sobre el proceso científico: naturaleza de la ciencia, investigación científica y relaciones ciencia-tecnología-sociedad, de carácter teórico y práctico. Se realizó una sesión teórica sobre naturaleza de la ciencia y dos prácticas: lectura de un artículo para aplicar los contenidos de la naturaleza de la ciencia y análisis de un proyecto realizado q...

Effect of spray on performance of the hydrogen mitigation system during LB-LOCA for CPR1000 NPP

International Nuclear Information System (INIS)

Huang, X.G.; Yang, Y.H.; Cheng, X.; Al-Hawshabi, N.H.A.; Casey, S.P.

2011-01-01

Highlights: → This paper presents the spray effect on HMS during LB-LOCA by using GASFLOW. → The positive and negative effects of spray are summarized. → And the combination of DIS and PAR system is suggested as reasonable countermeasures. → This research is an important work aimed at the study of spray and hydrogen mitigation. → The contents of this paper should become a required part of the safety analysis of Chinese NPPs. - Abstract: During the course of the hypothetical large break loss-of-coolant accident (LB-LOCA) in a nuclear power plant (NPP), hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel (RPV). It is then ejected from the break into the containment along with a large amount of steam. Management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen mitigation system (HMS) and spray system in CPR1000 NPP. The computational fluid dynamics (CFD) code GASFLOW is utilized in this study to analyze the spray effect on the performance of HMS during LB-LOCA. Results show that as a kind of HMS, deliberate igniter system (DIS) could initiate hydrogen combustion immediately after the flammability limit of the gas mixture has been reached. However, it will increase the temperature and pressure drastically. Operating the DIS under spray condition could result in hydrogen combustion being suppressed by suspended droplets inside the containment. Furthermore, the droplets could also mitigate local the temperature rise. Operation of a PAR system, another kind of HMS, consumes hydrogen steadily with a lower recombination rate which is not affected noticeably by the spray system. Numerical results indicate that the dual concept, namely the integrated application of DIS and PAR systems, is a constructive improvement for hydrogen safety under spray condition during LB-LOCA.

Methodology for LOCA analysis and its qualification procedures for PWR reload licensing

International Nuclear Information System (INIS)

Serrano, M.A.B.

1986-01-01

The methodology for LOCA analysis developed by FURNAS and its qualification procedure for PWR reload licensing are presented. Digital computer codes developed by NRC and published collectively as the WREM package were modified to get versions that comply to each requirement of Brazilian Licensing Criteria. This metodology is applied to Angra-1 basic case to conclude the qualification process. (Author) [pt

LOCA assessment experiments in a full-elevation, CANDU-typical test facility

International Nuclear Information System (INIS)

Ingham, P.J.; McGee, G.R.; Krishnan, V.S.

1990-01-01

The RD-14 thermal-hydraulics test facility, located at the Whiteshell Nuclear Research Establishment, is a full-elevation model representative of a CANDU primary heat transport system. The facility is scaled to accommodate a single, full-scale (5.0 MW, 21 kg/s), electrically heated channel per pass. The steam generators, pumps, headers, feeders and heated channels are arranged in a typical CANDU figure-of-eight geometry. The loop has an emergency coolant injection system (ECI) that may be operated in several modes, including typical features of the various ECI systems found in CANDU reactors. A series of experiments has been performed in RD-14 to investigate the thermal-hydraulic behaviour during the blowdown and injection phases of a loss-of-coolant accident (LOCA). The tests were designed to cover a full range of break sizes from feeder-sized breaks to guillotine breaks in either an inlet or an outlet header. Breaks resulting in channel flow stagnation were also investigated. This paper reviews the results of some of the LOCA tests carried out in RD-14, and discusses some of the behaviour observed. Plans for future experiments in a multiple-channel RD-14 facility, modified to contain multiple flow channels, are outlined. (orig.)

Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

International Nuclear Information System (INIS)

Ferreira Guimaraes, Antonio Cesar; Franklin Lapa, Celso Marcelo; Lamego Simoes Filho, Francisco Fernando; Cabral, Denise Cunha

2011-01-01

Research highlights: → This article presents an uncertainty modelling study using a fuzzy approach. → The AP1000 Westinghouse NPP was used and it is provided of passive safety systems. → The use of advanced passive safety systems in NPP has limited operational experience. → Failure rates and basic events probabilities used on the fault tree analysis. → Fuzzy uncertainty approach was employed to reliability of the AP1000 large LOCA. - Abstract: This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

La naturaleza y el ser humano como expresión de sentimientos

OpenAIRE

Pérez Rodríguez, Dámaris

2017-01-01

La psicología y los sentimientos están inmersos en mi trabajo y me gusta expresar en el lienzo o el papel la manera que tiene la mente de reaccionar frente a éstos, me fascina cómo el alma de una persona o de un espacio puede transmitir numerosas emociones, y todo esto toma forma de naturaleza. Universidad de Sevilla. Grado en Bellas Artes

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Science.gov (United States)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

Detection and control of potential core damage during a small-break LOCA

International Nuclear Information System (INIS)

Thomas, G.R.; Zebroski, E.L.

1981-01-01

A refreshing development in small-break LOCA analysis and testing is the recognition that this work can be of real value to a plant operator. Event-trees, or safety sequence diagrams, are being made increasingly realistic and are being used to develop and to test the abnormal transient operating guidelines (ATOGs) which provide a basis for operator response, training, and simulator work now under way. Perhaps the most monumental lesson of the TMI-2 accident is that the tradition of extreme worst-case analysis and data gathering can provide a directly negative contribution to safety if it is used as the basis for designing procedures and training of operator response. The event-trees for guiding operator response to abnormal conditions, ATOGs, must be based on physically realistic, best-estimate models. Possibly, the most dramatic risk reduction achieved will be attained through the use of realistic accident analysis, which leads to realistic operator guidelines and training, improved display of the critical information to the operator, and improved management structure. Given the dominant contribution of small-break LOCAs to the overall public risk envelope, realism should be the primary banner for future work in this field

PHEBUS program: first results on PWR fuel behaviour in LOCA conditions

International Nuclear Information System (INIS)

Del Negro, R.; Reocreux, M.; Pelce, J.; Legrand, B.; Berna, P.

1982-09-01

In the first PHEBUS test with pressurized rods some rods burst and clad temperature reached 1100 0 C in the 25 rods bundle. There is now a lot of valuable experimental results and their analysis is in progress. The phase II on fuel behaviour in case of a large LOCA will start at the beginning of 83. The onset of the SFD program is foreseen to take place on the first months of 85

Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code

International Nuclear Information System (INIS)

Perianez Alvarez, V.

2013-01-01

The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

Naturaleza, montaña, deporte y aventura en la vida de Santiago Ramón y Cajal

Directory of Open Access Journals (Sweden)

Eduardo Garrido Marín

2014-01-01

Full Text Available Santiago Ramón y Cajal fue un hombre excepcional. Su nombre estará siempre unido a la ciencia, pero su vida estuvo comprometida también con otras facetas que son, en general, muy poco conocidas. El deporte, la aventura, la montaña y la naturaleza jugaron un papel trascendental en forjar su personalidad, en el devenir de su salud y en la esencia de su destino científico por el cual es universalmente renombrado. El presente documento se centra en este hecho mediante el análisis del rico legado literario que nos dejó este sabio y, muy especialmente, en lo relacionado con su pasión por la montaña y la naturaleza.

Corporeidad y significado: naturaleza dual de lo físico

OpenAIRE

Arranz Rodrigo, Marceliano

2015-01-01

La información es una dimensión de lo real que, a pesar de no tener natu-raleza física, solo existe apoyada en una base física. Y su presencia en los objetos, eventos y procesos que constituyen nuestro entorno vital humano, es lo que permite que el mundo sea inteligible. Sin información no serían posibles las definiciones, ni el pensamiento, ni el lenguaje, ni la ciencia. La importancia de la información está creciendo de manera expo-nencial en nuestros días. Aclarar su estatuto ...

Utilizing elements of the CSAU phenomena identification and ranking table (PIRT) to qualify a PWR non-LOCA transients system code

Energy Technology Data Exchange (ETDEWEB)

Greene, K.R.; Fletcher, C.D.; Gottula, R.C.; Lindquist, T.R.; Stitt, B.D. [Framatome ANP, Richland, WA (United States)

2001-07-01

Licensing analyses of Nuclear Regulatory Commission (NRC) Standard Review Plan (SRP) Chapter 15 non-LOCA transients are an important part of establishing operational safety limits and design limits for nuclear power plants. The applied codes and methods are generally qualified using traditional methods of benchmarking and assessment, sample problems, and demonstration of conservatism. Rigorous formal methods for developing code and methodology have been created and applied to qualify realistic methods for Large Break Loss-of-Coolant Accidents (LBLOCA's). This methodology, Code Scaling, Applicability, and Uncertainty (CSAU), is a very demanding, resource intensive, process to apply. It would be challenging to apply a comprehensive and complete CSAU level of analysis, individually, to each of the more than 30 non-LOCA transients that comprise Chapter 15 events. However, certain elements of the process can be easily adapted to improve quality of the codes and methods used to analyze non- LOCA transients. One of these elements is the Phenomena Identification and Ranking Table (PIRT). This paper presents the results of an informally constructed PIRT that applies to non-LOCA transients for Pressurized Water Reactors (PWR's) of the Westinghouse and Combustion Engineering design. A group of experts in thermal-hydraulics and safety analysis identified and ranked the phenomena. To begin the process, the PIRT was initially performed individually by each expert. Then through group interaction and discussion, a consensus was reached on both the significant phenomena and the appropriate ranking. The paper also discusses using the PIRT as an aid to qualify a 'conservative' system code and methodology. Once agreement was obtained on the phenomena and ranking, the table was divided into six functional groups, by nature of the transients, along the same lines as Chapter 15. Then, assessment and disposition of the significant phenomena was performed. The PIRT and

Comparison of models discribing cladding deformations during LOCA

International Nuclear Information System (INIS)

Chakraborty, A.K.; Zipper, R.

1981-05-01

This report compares the important models for the determination of cladding deformations during LOCA. In addition to the comparisons of underlying assumptions of different models the same is done for the coefficients applied for the models. In order to assess the predictive capability of the models the calculated results are compared with the experimental results of the individual claddings. It was found out that the results of temperature ramp tests could be calculated better than that of the pressure ramp tests. The calculations revealed that even with the simplified assumption of the model used in TESPA the agreement of the calculated results with those of model NORA was relatively good. (orig.) [de

Sensitivity assessment of fuel performance codes for LOCA accident scenario

Energy Technology Data Exchange (ETDEWEB)

Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

2017-07-01

FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

Sensitivity assessment of fuel performance codes for LOCA accident scenario

International Nuclear Information System (INIS)

Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

2017-01-01

FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

Comparison of LOCA safety analysis in the USA, FRG, and Japan

International Nuclear Information System (INIS)

Leach, L.P.; Ybarrondo, L.J.; Hicken, E.F.; Tasaka, K.

1983-01-01

The bases for loss-of-coolant accident (LOCA) safety analysis required by reactor licensing regulations in the United States of America (USA), Federal Republic of Germany (FRG), and Japan are investigated and related to new data obtained since the regulations were established. The licensing approaches used in the three countries are similar in that a conservative calculation is called for, the necessary conservatism is unspecified, and new research data have had only limited effect on changing the regulations

UP601, a standardized botanical composition composed of Morus alba, Yerba mate and Magnolia officinalis for weight loss.

Science.gov (United States)

Yimam, Mesfin; Jiao, Ping; Hong, Mei; Brownell, Lidia; Lee, Young-Chul; Hyun, Eu-Jin; Kim, Hyun-Jin; Nam, Jeong-Bum; Kim, Mi-Ran; Jia, Qi

2017-02-16

The prevalence of obesity is surging in an alarming rate all over the world. Pharmaceutical drugs are considered potential adjunctive therapy to lifestyle modification. However, for most, besides being too expensive, their long term usages are hindered by their severe adverse effects. Here we describe the effect of UP601, a standardized blend of extracts from Morus alba, Yerba mate and Magnolia officinalis, in modulating a number of obesity-related phenotypic and biochemical markers in a high-fat high-fructose (HFF)-induced C57BL/6J mouse model of obesity. Adipogenesis activity of the composition was assessed in 3T3-L1 cells in vitro. Effects of UP601 on body weight and metabolic markers were evaluated. It was administered at oral doses of 300 mg/kg, 450 mg/kg and 600 mg/kg for 7 weeks. Orlistat (40 mg/kg/day) was used as a positive control. Body compositions of mice were assessed using dual energy X-ray absorptiometry (DEXA). Serum biomarkers were measured for liver function and lipid profiling. Relative organ weights were determined. Histopathological analysis was performed for non-alcoholic steatohepatitis (NASH) scoring. UP601 at 250 μg/ml resulted in 1.8-fold increase in lipolysis. Statistically significant changes in body weight (decreased by 9.1, 19.6 and 25.6% compared to the HFF group at week-7) were observed for mice treated with UP601 at 300, 450 and 600 mg/kg, respectively. Reductions of 9.1, 16.9, and 18.6% in total cholesterol; 45.0, 55.0, 63.6% in triglyceride; 34.8, 37.1 and 41.6% in LDL; 3.2, 21.6 (P = 0.03) and 33.7% (P = 0.005) in serum glucose were observed for UP601 at 300, 450 and 600 mg/kg, respectively. Body fat distribution was found reduced by 31.6 and 17.2% for the 450 mg/kg UP601 and orlistat, respectively, from the DEXA scan analysis. Up to an 89.1% reduction in mesenteric fat deposit was observed for UP601 in relative organ weight. Statistically significant improvements in NASH scores were observed for mice treated

LOCA testing of high burnup PWR fuel in the HBWR. Additional PIE on the cladding of the segment 650-5

Energy Technology Data Exchange (ETDEWEB)

Oberlaender, B.C.; Espeland, M.; Jenssen, H.K.

2008-07-01

IFA-650.5, a test with pre-irradiated fuel in the Halden Project LOCA test series, was conducted on October 23rd, 2006. The fuel rod had been used in a commercial PWR and had a high burnup, 83 MWd/kgU. Experimental arrangements of the fifth test were similar to the preceding LOCA tests. The peak cladding temperature (PCT) level was higher than in the third and fourth tests, 1050 C. A peak temperature close to the target was achieved and cladding burst occurred at approx. 750 C. Within the joint programme framework of the Halden Project PIE was done, consisting of gamma scanning, visual inspection, neutron-radiography, hydrogen analysis and metallography / ceramography. An additional extensive PIE including metallography, hydrogen analysis, and hardness measurements of cross-sections at seven axial elevations was done. It was completed to study the high burnup and LOCA induced effects on the Zr-4 cladding, namely the migration of oxygen into the cladding from the inside surface, the cladding distension, and the burst (author)(tk)

El poder y la fuerza. Apuntes doctrinales sobre la naturaleza revolucionaria de las FARC.

Directory of Open Access Journals (Sweden)

Vicente Torrijos R.

2004-01-01

Full Text Available Con base en su experiencia como investigador del conflicto aramdo colombiano, el autor presenta un artículo de reflexión acerca de la naturaleza de las Farc, principal organizaciòn armada de Colombia y del hemisferio occidental, y se analizan sus capacidades para desestabilizar el sistema democrático nacional.

Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

Energy Technology Data Exchange (ETDEWEB)

Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

2015-11-15

Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

BWR 1 % main recirculation line break LOCA tests, RUNs 917 and 918, without HPCS at ROSA-III program

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Okazaki, Motoaki; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

1988-07-01

In a case of small break loss-of-coolant accident (LOCA) at a boiling water reactor (BWR) system, it is important to lower the system pressure to cool down the reactor system by using either the high pressure core spray (HPCS) or the automatic depressurization system (ADS). The report presents characteristic test results of RUNs 918 and 917, which were performed at the rig-of-safety assessment (ROSA)-III program simulating a 1 % break BWR LOCA with an assumption of HPCS failure, and clarifies effects of the ADS delay time on a small break LOCA. The ROSA-III test facility simulates principal components of a BWR/6 system with volumetric scaling factor of 1/424. It is experimentally concluded that the ADS delay time shorter than 4 minutes results in a similar PCT as that in a standard case, in which the PCT is observed after actuation of the low pressure core spray (LPCS). And the ADS delay time longer than 4 minutes results in higher PCT than in the standard case. In the latter, the PCT depends on the ADS time, a 220 K higher PCT, for example, in a case of 10 minutes ADS delay compared with the standard case. (author) 52 refs. 299 figs

Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

International Nuclear Information System (INIS)

Urbonavičius, E.; Povilaitis, M.; Kontautas, A.

2015-01-01

Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

Benchmark Tests to Develop Analytical Time-Temperature Limit for HANA-6 Cladding for Compliance with New LOCA Criteria

Energy Technology Data Exchange (ETDEWEB)

Lee, Sung Yong; Jang, Hun; Lim, Jea Young; Kim, Dae Il; Kim, Yoon Ho; Mok, Yong Kyoon [KEPCO Nuclear Fuel Co. Ltd., Daejeon (Korea, Republic of)

2016-10-15

According to 10CFR50.46c, two analytical time and temperature limits for breakaway oxidation and postquench ductility (PQD) should be determined by approved experimental procedure as described in NRC Regulatory Guide (RG) 1.222 and 1.223. According to RG 1.222 and 1.223, rigorous qualification requirements for test system are required, such as thermal and weight gain benchmarks. In order to meet these requirements, KEPCO NF has developed the new special facility to evaluate LOCA performance of zirconium alloy cladding. In this paper, qualification results for test facility and HT oxidation model for HANA-6 are summarized. The results of thermal benchmark tests of LOCA HT oxidation tester is summarized as follows. 1. The best estimate HT oxidation model of HANA- 6 was developed for the vender proprietary HT oxidation model. 2. In accordance with the RG 1.222 and 1.223, Benchmark tests were performed by using LOCA HT oxidation tester 3. The maximum axial and circumferential temperature difference are ± 9 .deg. C and ± 2 .deg. C at 1200 .deg. C, respectively. At the other temperature conditions, temperature difference is less than 1200 .deg. C result. Thermal benchmark test results meet the requirements of NRC RG 1.222 and 1.223.

A simple analytical scaling method for a scaled-down test facility simulating SB-LOCAs in a passive PWR

International Nuclear Information System (INIS)

Lee, Sang Il

1992-02-01

A Simple analytical scaling method is developed for a scaled-down test facility simulating SB-LOCAs in a passive PWR. The whole scenario of a SB-LOCA is divided into two phases on the basis of the pressure trend ; depressurization phase and pot-boiling phase. The pressure and the core mixture level are selected as the most critical parameters to be preserved between the prototype and the scaled-down model. In each phase the high important phenomena having the influence on the critical parameters are identified and the scaling parameters governing the high important phenomena are generated by the present method. To validate the model used, Marviken CFT and 336 rod bundle experiment are simulated. The models overpredict both the pressure and two phase mixture level, but it shows agreement at least qualitatively with experimental results. In order to validate whether the scaled-down model well represents the important phenomena, we simulate the nondimensional pressure response of a cold-leg 4-inch break transient for AP-600 and the scaled-down model. The results of the present method are in excellent agreement with those of AP-600. It can be concluded that the present method is suitable for scaling the test facility simulating SB-LOCAs in a passive PWR

Primary LOCA in VVER-1000 by pressurizer PORV failure

International Nuclear Information System (INIS)

Sabotinov, L.; Lutsanych, S.; Kadenko, I.

2013-01-01

The paper presents the calculations and analysis of the design basis accident of a standard VVER-1000/V320 reactor with inadvertent opening and stuck in open position of the pressurizer pilot operated relief valve (PORV). The objective is the independent assessment of this accident applying the French best estimate thermal-hydraulic computer code CATHARE 2 and verification to meet the safety criteria for such kind of the accident. The 'Inadvertent opening and stuck in open position of PORV' is a design basis accident classified as Medium Break Loss of Coolant Accident (MB LOCA) with the equivalent diameter of the break D- 68 mm. This accident is particularly interesting to be calculated and analyzed, because it took place at operating NPP with VVER-1000 reactors (Rovno NPP) in 2009. The calculations have been carried out with conservative conditions as usual for DBA analysis. The NPP model corresponds to a real VVER-1000/V320 configuration and comprises all safety systems, adopted for one of the latest CATHARE 2 versions. The results of CATHARE 2 calculations are compared with available results of RELAP5 calculations. There is similarity of the thermal-hydraulic parameters behavior, but also some differences can be observed basically due to the break flow prediction, which causes differences in primary pressure evaluation. Both calculations show that there is no boiling crisis in the reactor core and reliable cooldown is achieved. The calculations performed with CATHARE2 code demonstrate the ability of the code to predict reasonably the break flow, pressures, temperatures etc. for considered LOCA scenario and to be applied for safety studies

LOCA Analysis of KAIST-Micro Modular Reactor with Modified GAMMA+ code

Energy Technology Data Exchange (ETDEWEB)

Oh, Bong Seong; Ahn, Yoon Han; Kim, Seong Gu; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

The supercritical carbon dioxide (S-CO{sub 2}) power cycle is being seriously investigated around the world due to its simple layout, quite high efficiency around 500 .deg. C turbine inlet temperature, etc. By combining these two ideas, the KAIST research team developed a S-CO{sub 2} cooled SMR, called KAIST-Micro Modular reactor (MMR), which is targeting transportability and electricity supply for remote region. Therefore, requirements of MMR design are factory fabrication of the total system including power conversion system to be transported and air cooling to be independent from the site selection. Until now, steady performances and sizes of components were evaluated. Thus, in this paper a transient performance of the MMR are simulated with special focus on the loss of coolant accident (LOCA) at cold leg pipe. The MMR is a newly suggested innovative small modular reactor concept by the KAIST research team. Since the MMR is cooled by supercritical CO{sub 2}, general safety codes for conventional reactors have limitations. Thus, GAMMA+ code for the transient analysis of a gas-cooled reactor was selected and modified for the S-CO{sub 2} power system. After the modification of GAMMA+ code, LOCA is simulated, which is considered as one of the most limiting accidents in terms of safety of nuclear power plant.

Impact of 2D/3D-project on LOCA-licensing analysis and reactor safety of PWRs

International Nuclear Information System (INIS)

Winkler, F.; Krebs, W.D.

1989-01-01

In the past LOCA-licensing analysis has included large conservatisms to compensate for the lack of detailed two phase flow and full scale experimental data. The 2D/3D-project was established to improve the data base in order to minimize the conservatisms required. The significant results and findings of the full scale Upper Plenum Test Facility (UPTF) and from the electrically heated Slab Core Test Facility (SCTF) were particularly useful for understanding the multidimensional phenomena in the primary system and in the core of a PWR. UPTF results were used to verify the TRAC-PF1 analysis of a PWR with combined ECC-Injection during the reflood phase of a large break-LOCA. Comparison of these results with results from classic licensing calculations quantifies the large safety margin in earlier licensing procedures and in reactor systems. (orig.)

The Effect of Mg2+, Cu2+ and Zn2+ pre-treatment on the color of yerba maté (Ilex paraguariensis leaves

Directory of Open Access Journals (Sweden)

Griselda Patricia Scipioni

2010-12-01

Full Text Available The aim of this work was to study the effect of alkaline blanching followed by an immersion in salt solutions of Mg2+, Cu2+ and Zn2+ on color preservation of Yerba Maté leaves. The concentration of NaOH in the alkaline solution influenced all of the color parameters. Ion concentration and dipped time influenced only some color parameters. The color parameters of the product obtained with different treatments were different from the control sample (blanched with boiling water instead of alkaline blanching and from the product obtained via industrial processing. The green color in the pre-treated and control samples was more intense (greater values of -a and darker (low values of L and b. Ionic and ash content in the leaves increased with the treatments.

La enseñanza de las Ciencias de la Naturaleza en Educación Infantil mediante el trabajo por proyectos

OpenAIRE

Herreros Rodríguez, Lucía

2015-01-01

Este trabajo, es un proyecto educativo para la enseñanza de las Ciencias de la Naturaleza. Se basa en la metodología por proyectos en la etapa de educación infantil. En el se aborda la importancia de este innovador método de enseñanza/aprendizaje. En educación infantil, es esencial el estudio de las Ciencias de la Naturaleza, ya que le permite a los niños comprender la realidad más cercana que le rodea. En la propuesta que se plantea, se integran contenidos de las diferentes áreas y diferente...

El nuevo paradigma genético y la naturaleza humana: Una perspectiva desde la bioética reflexiva y secular

OpenAIRE

Orlando, Mejía

2005-01-01

La genética se ha convertido en el área científica con mayores implicaciones bioéticas y sociales, debido al poder de transformación que tiene sobre lo que se ha considerado, hasta ahora, como la naturaleza humana. En este trabajo se propone analizar la relación entre el nuevo paradigma genético, la noción histórica de la esencia humana y los contemporáneos modelos antropocéntrico y evolutivo de la naturaleza humana. Utilizando elementos conceptuales que correspondan a una bioética reflexiva,...

Lumped-parameter modeling of PWR downcomer and pressurizer for LOCA conditions

International Nuclear Information System (INIS)

Rohatgi, U.S.; Saha, P.; Dubow, A.A.

1978-01-01

Two lumped-parameter models, one for a PWR downcomer and the other for a pressurizer, are presented. The models are based on the transient, nonhomogeneous, drift-flux description of two-phase flow, and are suitable for simulating a hypothetical LOCA condition. Effects of thermal nonequilibrium are incorporated in the downcomer model, whereas the pressurizer model can track the interfaces among various flow regimes. Semiimplicit numerical schemes are used for solution. Encouraging results have been obtained for both the models. (author)

Belleza sin autoría: una visión pluralista de la apreciación estética de la naturaleza

OpenAIRE

Arribas Herguedas, Fernando

2015-01-01

¿Puede la naturaleza ser estéticamente apreciada tal y como lo son las obras de arte? ¿Es posible establecer criterios objetivos para apreciar correctamente la naturaleza? ¿Cuáles son los rasgos que definen una adecuada apreciación estética de los objetos y paisajes naturales? El debate sobre estas cuestiones debe gran parte de su importancia al “cognitivismo estético” del filósofo Allen Carlson. Según este enfoque, el conocimiento científico de los objetos y entornos naturales representa par...

Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

Energy Technology Data Exchange (ETDEWEB)

Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

2008-10-15

A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

Core heat transfer analysis during a BWR LOCA simulation experiment at ROSA-III

International Nuclear Information System (INIS)

Yonomoto, T.; Koizumi, Y.; Tasaka, K.

1987-01-01

The ROSA-III test facility is a 1/424-th volumetrically scaled BWR/6 simulator with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA). Heat transfer analyses for 5, 15, 50 and 200% break tests were conducted to understand the basic heat transfer behavior in the core under BWR LOCA conditions and to obtain a data base of post-critical heat flux (CHF) heat transfer coefficients and quench temperature. The results show that the convective heat transfer coefficient of dried-out rods at the core midplane during a steam cooling period is less than approximately 120 W/m 2 K. It is larger than existing data measured at lower pressures during a spray cooling period. Bottom-up quench temperatures are given by a simple equations: The sum of the saturation temperature and a constant of 262 K. Then the heat transfer model in the RELAP4/MOD6/U4/J3 code was revised using the present results. The rod surface temperature behavior in the 200% break test was calculated better by using the revised model although the model is very simple. (orig.)

Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

Energy Technology Data Exchange (ETDEWEB)

Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

Energy Technology Data Exchange (ETDEWEB)

Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

2017-06-01

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

LOFA [loss of flow accident] and LOCA [loss of coolant accident] in the TIBER-II engineering test reactor: Appendix A-4

International Nuclear Information System (INIS)

Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

1987-01-01

This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510 0 C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs

ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

Energy Technology Data Exchange (ETDEWEB)

Blomquist, R; Shen, K; Sjoeberg, A

1995-03-01

This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs.

ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

International Nuclear Information System (INIS)

Blomquist, R.; Shen, K.; Sjoeberg, A.

1995-03-01

This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs

Fitness for service after a LOCA: A process applied to Pickering NGS Unit 2

International Nuclear Information System (INIS)

McLean, J.A.; Beaton, D.L.

1996-01-01

The fitness for service process provides a unique proven methodology for assessing and correcting post-LOCA damage, essential to plant restart. The process uses the as-built plant configuration for modelling input and features self correcting feedback from inspection to validate assessment models. This paper focuses on the process steps and the infrastructure necessary to execute the process

Uncertainties in radioactivity release from LWR plants under LOCA conditions - magnitude and consequences

International Nuclear Information System (INIS)

Mattila, L.J.

1977-01-01

Standardized, deterministic, and supposedly conservative calculation methods and parameter values are applied in radiological safety analyses required for licensing individual nuclear power plants. As realistic as possible and comprehensive analyses are, however, absolutely necessary for many purposes, such as developing improved designs, comparisons between nuclear and non-nuclear power plant alternatives or entire energy production strategies, and also formulating improved acceptance criteria for plant licensing. A specific type of LOCA, called design basis accident (DBA), has obtained an exceptionally important status in the licensing procedure of light water reactor nuclear power plants. This postulated accident has a decisive influence on plant siting and on the design of the various engineered safety features. To avoid certain potential negative effects of the highly standardized guideline-based accident analysis procedure - such as introduction of apparent design ''improvements'', wrong priorization of research efforts, etc. - and to provide a realistic view about the safety of light water reactors to supplement the conservative results from regulatory analyses, a comprehensive understanding of the radiological consequences of LOCA's is indispensable. Estimates of fission product release from LWR plants under different LOCA conditions are associated with uncertainties due to deficient knowledge and truly random variability. The following steps of the fission product transport chain are discussed: generation of activity, fission product release from fuel to fuel pin voids prior to the accident, fuel rod puncturing and fission product release from punctured rods during the accident, further release from fuel during the transient, transport to the containment and finally removal in and leakage from the containment. Numerical examples are given by comparing assumptions, parameter values, and results from the following four analyses: the present guideline

Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

International Nuclear Information System (INIS)

Fischer, S.R.; Farman, R.F.; Birdsell, S.A.

1992-01-01

This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model

Conexiones naturales de la gran ciudad con las áreas periurbanas de Naturaleza. Aplicación al caso de Madrid

Directory of Open Access Journals (Sweden)

Alicia Bedmar Perlado

2015-09-01

Full Text Available La movilidad peatonal se está convirtiendo cada vez más en un en objeto de atención para la sociedad. Después de un largo periodo de abandono debido al auge del transporte privado y al deterioro consecuente del espacio público, se mira al peatón como un agente regenerador de la ciudad, capaz de promover la integración social. Este hecho y la necesidad de acercar la Naturaleza a las ciudades motivan la búsqueda de soluciones para conseguir este fin.Se probará una metodología basada en estudios recientes sobre movilidad peatonal, para posibilitar la llegada de los ciudadanos a las áreas de Naturaleza cercanas. Para probar y refinar la metodología se aplica a un estudio de caso, la ciudad de Madrid. El análisis de las posibles salidas de la ciudad, contrastará las conclusiones sobre la marcha a pie dentro de la ciudad. Como conclusión del estudio de caso se analizará la accesibilidad desde los distritos de Madrid a los puntos de Naturaleza cercanos.

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

International Nuclear Information System (INIS)

Khvostov, G.; Wiesenack, W.; Zimmermann, M.A.; Ledergerber, G.

2011-01-01

Highlights: → A model for the dynamics of axial gas redistribution in fuel rods during the LOCA is developed and coupled to the FALCON fuel behaviour code. → The first verification of the model is carried out using the data of the selected Halden LOCA tests. → According to calculation, the short rods used in the Halden tests show a small effect of the delayed gas redistribution during the clad ballooning. → The predicted effect is significant in the full length rods, eventually resulting in a considerable delay of the predicted moment of cladding rupture. → The predicted delay of cladding burst may be large enough to eventually affect the efficiency of the emergency core cooling system. - Abstract: A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code. The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from

Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

Energy Technology Data Exchange (ETDEWEB)

De, T K; Collins, W M; Holmes, R W [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

CANDU nuclear reactors use D{sub 2}0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs.

Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

International Nuclear Information System (INIS)

De, T.K.; Collins, W.M.; Holmes, R.W.

1995-01-01

CANDU nuclear reactors use D 2 0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs

TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

International Nuclear Information System (INIS)

MAJI, A. K.; MARSHALL, B.

2000-01-01

In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

Parte de la Naturaleza: sobre el cuerpo como potencia compositiva

Directory of Open Access Journals (Sweden)

Juliana Merçon

2012-11-01

Full Text Available http://dx.doi.org/10.5007/2175-795X.2012v30n2p473 En medio de la constelación de mitos que han contribuido a la actual crisis socioecológica,se encuentra la creencia de que somos individuos sustanciales, atomizadoso separados del entorno al cual intentamos ‘dominar’. La ignorancia con respecto anuestra interdependencia ontológica está en la base de muchas de nuestras posturas(auto destructivas y ha sido reforzada por diversos sistemas teóricos a lo largo de lahistoria occidental. Al margen de dualismos platónicos, de transcendencias judaicocristianasy del cogito cartesiano, la filosofía de Benedictus de Spinoza (1632-1677nos invita a repensar los fundamentos de la realidad compleja en la cual estamosinmersos. Mi propósito en este artículo es tejer relaciones productivas entre algunasdimensiones de su filosofía, el pensamiento ambiental y la educación. El punto departida será la concepción spinozista de cuerpo. Este será definido como potencia quese expresa a la vez como productividad (fuerza causal y sensibilidad (fuerza receptiva.El carácter esencialmente relacional y afectivo del cuerpo humano nos permitirápensarlo también como potencia compositiva. Desconstruyendo la dicotomíaentre Naturaleza y Cultura, propondré que reflexionemos sobre las composicioneshumano-tecnológicas con respecto a sus efectos sobre nuestra potencia de pensar yactuar. En conclusión, presentaré algunos comentarios sobre el rol de la educacióncomo espacio relacional, en el cual se pueden integrar de forma potenciadora lasdimensiones epistémica, ética, política y ambiental de nuestra existencia como partesinterdependientes de la Naturaleza.

Best estimate modeling of fuel thermomechanical behaviour in WWER 1000 LB LOCA

International Nuclear Information System (INIS)

Valach, M.; Klouzal, J.; Zymak, J.; Dostal, M.

2009-01-01

The paper summarizes our calculations of the performance of the WWER 1000 NPP fuel rods during postulated LB LOCA. The thermomechanical modeling was performed by FRAPTRAN using the FRACAS-I mechanical model using the boundary conditions calculated by the ATHLET code. The results and their statistical evaluation are presented, the process of the generalization of gained insight into the best-estimate thermal-hydraulic analyses (BE TM) predictions in order to define a generic BE TM methodology is outlined (authors)

IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU

International Nuclear Information System (INIS)

Cunningham, Mitchel E.; Turnbull, J.A.

2003-01-01

Description - Objectives - Results: The U.S. Nuclear Regulatory Commission (NRC) conducted a series of thermal-hydraulic and cladding mechanical deformation tests in the National Research Universal (NRU) reactor at the Chalk River National Laboratory in Canada. The objective of these tests was to perform simulated loss-of-coolant-accident (LOCA) experiments using full-length light-water reactor fuel rods to study mechanical deformation, flow blockage, and coolability. Three phases of a LOCA (i.e., heat-up, reflood, and quench) were performed in situ using nuclear fissioning to simulate the low-level decay power during a LOCA after shutdown. All tests used PWR-type, non-irradiated fuel rods. Provided here is information on two materials tests, MT-6A and MT-4, which PNNL considers the better characterized for the purposes of setting up computer cases. The NRU reactor is a heterogeneous, thermal, tank-type research reactor. It has a power level of 135 MWth and is heavy-water moderated and cooled. The coolant has an inlet temperature of 310 K at a pressure of 0.65 MPa. The MT tests were conducted in a specially designed test train to supply the specified coolant conditions of flowing steam, stagnant steam, and then reflood. Typical instrumentation for the MT tests included fuel centerline thermocouples, cladding inner surface thermocouples, cladding outer surface thermocouples, rod internal gas pressure transducers or pressure switches, coolant channel steam probes, and self-powered neutron detectors. This instrumentation allowed for determining rupture times and cladding temperature. The test rods for the LOCA cases in the NRU reactor were irradiated in flowing steam prior to the transient, stagnant steam during the transient and prior to reflood, and then reflood conditions to complete the transient. Both cladding inner surface and outer surface temperatures were measured, in addition to coolant temperatures. However, only cladding inner surface temperatures were

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

Energy Technology Data Exchange (ETDEWEB)

Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

Interés por el gobierno y gobierno a través del interés: constitución de la naturaleza infantil

Directory of Open Access Journals (Sweden)

Dora Lilia Marín-Díaz

2013-05-01

Full Text Available El  artículo  propone  una  lectura  histórica  de  la  forma como fue creada discursivamente la naturaleza infan-til.  Se señala  que fue a  través de la  invención y uso de nociones como desarrollo, interés, libertad, entre otras, que esa naturaleza infantil emergió y se plantea que tal intervención expresa la articulación de tres formas de practicar y pensar la vida en las sociedades occidentales –discursos liberales, discursos naturalistas y prácticas disciplinarias.  La  articulación  de  esos  tres  elementos aconteció durante lo que se conoce como Modernidad, momento en el que se organizaron formas de gobierno vinculadas a la idea de libertad. Palabras claveAprendizaje,  naturaleza  infantil,  gobierno.

Results of small break LOCA analysis for Kuosheng nuclear power plant using the RELAP5YA computer code

International Nuclear Information System (INIS)

Wang, L.C.; Jeng, S.C.; Chung, N.M.

2004-01-01

One lesson learned from the Three Mile Island (TMI) accident was the analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or nuclear fuel suppliers for small break Loss Of Coolant Accident (LOCA) analysis for compliance with appendix K to 10CFR50 should be revised, documented and submitted for USNRC approval and the plant-specific calculations using NRC-approved models for small-break LOCA to show compliance with 10CFR50.46 should be submitted for NRC approval. A study by Taiwan Power Company (TPC) under the guidance of Yankee Atomic Electric Company (YAEC) has been undertaken to perform this analysis for Kuosheng nuclear power plant. This paper presents the results of the analysis that are useful in satisfying the same requirements of the Republic Of China Atomic Energy Commission (ROCAEC). (author)

Prediction of moderator temperature under 35% RIH break LOCA with LOECC in CANDU calandria vessel

International Nuclear Information System (INIS)

Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung; Lee, Jae Yung

2004-01-01

A CANDU reactor has the unique safety features with the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors such as a PWR. One of the safety features is that the heavy water moderator is continuously cooled, providing with a heat sink for the decay heat produced in the fuel when there is the LOCA with the coincident failure of the emergency coolant injection (ECI) system. Under such a dual failure condition, the hot pressure tube (PT) would deform into contacting with the calandria tube (CT), providing with an effective heat transfer path from the fuel to the moderator. Following PT/CT contact, there is the spike of the heat flux in the moderator surrounding the CT, which could lead to sustained CT dryout. The prevention of the CT dryout depends on available local moderator subcooling. Higher moderator temperature (or lower subcooling) would decrease the margin of the CTs to dryout. As for LOCAs with coincident loss of the ECI, fuel channel integrity depends on the capability of the moderator as an ultimate heat sink. In this regard, the Canadian Nuclear Safety Commission (CNSC) had categorized the temperature prediction for the moderator cooling integrity as a general action item (GAI) and had recommended that a series of experimental works should be performed to verify the evaluation codes comparing with the results of three-dimensional experimental data. However, although a couple of computer codes were used to predict moderator temperature prediction for those problems, they could not be adequately validated due to the uncertainty of temperature prediction. In this work, the temperature prediction under the transient condition of LOCA with loss of emergency core cooling (LOECC) in a CANDU reactor is conducted using the optimized calculation scheme from the previous work

Naturaleza y función de comunicación en el marketing

OpenAIRE

Santillan, Paco

2007-01-01

En la actualidad vemos que el marketing necesita de la comunicación por lo que he creído necesario realizar una investigación sobre la naturaleza y función de la comunicación de Marketing. La comunicación es necesaria para mejorar el ajuste entre la oferta y demanda, alertando al mercado sobre la existencia y características de los bienes y servicios ofertados, como función complementaria de la de intercambio físico de tales bienes. La comunicación en marketing sirve pa...

Analysis of a large-break LOCA at lower operational modes

Energy Technology Data Exchange (ETDEWEB)

Jeong, H.Y.; Jun, H.Y.; Lee, K. [Korea Electric Power Corporation, Taejon (Korea)

2000-10-01

To improve Technical Specifications and Emergency Operating Guidelines (EOGs) applicable at lower operational modes it is required to perform the safety analysis reflecting the operational characteristics in those modes. Because the component availability and system configurations at lower modes are different from those of power mode, the plant safety at lower modes should be confirmed through independent analyses. In the present study, a large-break loss-of-coolant accident is analyzed to evaluate the containment pressure and temperature control function for the preparation of EOGs applicable at lower modes. To reach the required shutdown condition, the plant cool-down is controlled by the secondary steam flow and auxiliary feedwater. The mass and energy releases from primary system are obtained from RELAP5/MOD3.1 calculation and the containment pressure and temperature are evaluated with CONTEMPT-LT code. The reference plant is Korean Next Generation Reactor having 4,000 MW thermal power. Two cases of cold leg LOCA initiated at Mode 3 with and without SIT operation are calculated. At the given plant conditions, all safety injection pumps are still available. The calculation at the condition of maximum mass and energy release shows that the containment pressure and temperature can be controlled within acceptable criteria, which means the operations of 2 or 4 fan coolers are the possible success paths to achieve the containment P/T control safety function. The peak cladding temperature with minimum safety injection flow does not show remarkable excursion, which implies the lower mode LOCA at Mode 3 can be bounded by the results obtained at full power from the viewpoint of ECCS performance. (author)

Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

International Nuclear Information System (INIS)

Vigil, R.A.; Jacobus, M.J.

1994-04-01

Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent

Severe damage analysis of VVER 1000 following large break LOCA using Astec code

International Nuclear Information System (INIS)

Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

2007-01-01

Severe accident analysis of a reactor is an important aspect in the evaluation of source term. This in turn helps in emergency planning. An analysis has been carried out for VVER-1000 (V320) reactor following Large Break LOCA (loss of coolant accident) along with Station Blackout (SBO). Computer code ASTEC (jointly developed by IRSN, France, and GRS, Germany) is used for analyzing the transient. This integral code has been designed to be used as reference code for PSA2 studies. Severe accident analysis is carried out for an accident initiated by Large break LOCA along with SBO. Two cases have been analysed with the version ASTEC V1.2-rev1. In the first case hydro-accumulators are considered not available while the second case has been analysed with hydro accumulators. In this paper, ASTEC predictions have been studied for the in-vessel phase of the accident till vessel failure. The vessel failure was observed at 6979 s when accumulators were assumed not available. The vessel failure was quite delayed (19294 s) with operating accumulators. The hydrogen production was found to be very large (22% of total Zr inventory) in the case with accumulators compared to the case without accumulators (1.5% of total Zr inventory)

Space-time neutronic analysis of postulated LOCA's in CANDU reactors

International Nuclear Information System (INIS)

Luxat, J.C.; Frescura, G.M.

1978-01-01

Space-time neutronic behaviour of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated LOCA'S. The approach involves the efficient use of a set of computer codes which provide a capability to perform simulations ranging from detailed, accurate 3-dimensional space-time to low-cost survey calculations using point kinetics with some ''effective'' spatial content. A new, space-time kinetics code based upon a modal expansion approach is described. This code provides an inexpensive and relatively accurate scoping tool for detailed 3-dimensional space-time simulations. (author)

Evaluation on the habitability of a reactor control room for a 1300 MWe PWR following a LOCA

International Nuclear Information System (INIS)

Chang, Si Young; Ha, Chung Woo

1988-01-01

An evaluation on the habitability of a reactor control room for a French 1300 MWe P'4 type PWR following a LOCA has been performed through exposure dose assessment for a reactor operator. A computer code COREX calculating the time-integrated exposure dose has been developed to provide a reasonable basis in this evaluation. Using COREX the exposure dose reduction factors in the reactor control room, the time--integrated radioactivities released into the atmosphere and the time-integrated exposure dose up to 30 days following the LOCA can be also calculated. From the exposure dose assessment, the time-integrated exposure dose to whole body and thyroid of a reactor operator were 0.36 mSv(0.036 rem) and 480 mSv(48.0 rem), respectively after 30 days following the LOCA. The thyroid dose of 480 mSv was nearly 10 times greater than the dose equivalent limit of 50 mSv(5.0 rem) set by the ICRP. Regarding the habitability of a reactor control room, this exceeding thyroid exposure dose could be reduced to 1.2 mSv(0.12 rem), which is 400 times less than the original, by considering the practical 4 work-shifts a day, and by improving the iodine removal efficiency of the filtration system n the reactor control room through the reinforcement of charcoal bed filters for iodine removal. The radiological habitability of a reactor control room, therefore, could be assured by comparing with the dose equivalent limit of the ICRP

Notes on the Implementation of Non-Parametric Statistics within the Westinghouse Realistic Large Break LOCA Evaluation Model (ASTRUM)

International Nuclear Information System (INIS)

Frepoli, Cesare; Oriani, Luca

2006-01-01

In recent years, non-parametric or order statistics methods have been widely used to assess the impact of the uncertainties within Best-Estimate LOCA evaluation models. The bounding of the uncertainties is achieved with a direct Monte Carlo sampling of the uncertainty attributes, with the minimum trial number selected to 'stabilize' the estimation of the critical output values (peak cladding temperature (PCT), local maximum oxidation (LMO), and core-wide oxidation (CWO A non-parametric order statistics uncertainty analysis was recently implemented within the Westinghouse Realistic Large Break LOCA evaluation model, also referred to as 'Automated Statistical Treatment of Uncertainty Method' (ASTRUM). The implementation or interpretation of order statistics in safety analysis is not fully consistent within the industry. This has led to an extensive public debate among regulators and researchers which can be found in the open literature. The USNRC-approved Westinghouse method follows a rigorous implementation of the order statistics theory, which leads to the execution of 124 simulations within a Large Break LOCA analysis. This is a solid approach which guarantees that a bounding value (at 95% probability) of the 95 th percentile for each of the three 10 CFR 50.46 ECCS design acceptance criteria (PCT, LMO and CWO) is obtained. The objective of this paper is to provide additional insights on the ASTRUM statistical approach, with a more in-depth analysis of pros and cons of the order statistics and of the Westinghouse approach in the implementation of this statistical methodology. (authors)

Frequency probabilistic analysis of a small break LOCA due to a power operated relief valve (PORV) for Angra-1 pre-TMI and post-TMI

International Nuclear Information System (INIS)

Onusic Junior, J.

1986-01-01

After the TMI event efforts were aimed towards improvements in the operational and administrative procedures related to the power operated relief valves (PORVs) in order to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by stuck-open power operated relief valve. This paper presents a frequency probabilistic analysis of a small break LOCA due to a stuck open PORV and safety valve to the Angra I nuclear power plant in operating conditions pre-TMI and post-TMI. (Author) [pt

Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

Diversas concepciones en torno a la naturaleza como sujeto político. De la necesidad de cambio de paradigmas

Directory of Open Access Journals (Sweden)

José de Jesús Herrera Ospina

2015-08-01

Full Text Available RESUMEN: El presente texto relacionan perspectivas en torno al discurso que sobre la Naturaleza se han construido, haciendo especial énfasis en la necesidad de reconocerle (a la naturaleza la categoría de sujeto, en oposición a la herencia moderna, industrial, post industrial, que la considera un mero objeto, materia prima, insumo base para el sistema de producción en el marco del sistema capitalista. The present paper relates perspectives around the discourse that about nature have been built, with special emphasis on the need to recognize it (nature the category of subject, as opposed to the modern, industrial, post-industrial inheritance, which considers it to be a mere object, raw material, base input for the production system within the framework of the capitalist system.

LA SENDA BIOCÉNTRICA: VALORES INTRÍNSECOS, DERECHOS DE LA NATURALEZA Y JUSTICIA ECOLÓGICA

Directory of Open Access Journals (Sweden)

Eduardo Gudynas

2010-01-01

Full Text Available En este artículo se revisan las principales perspectivas conceptuales y prácticas sociales y políticas que defienden a la Naturaleza como sujeto de derechos, en contraste con las posturas convencionales que la entienden únicamente como objeto de valoración por los seres humanos. Se analizan los aportes sobre los valores intrínsecos en el ambiente, su expresión en posturas biocéntricas y los contrastes con el antropocentrismo propio de la Modernidad. Se consideran sus expresiones concretas en América Latina, especialmente en la nueva Constitución de Ecuador. Seguidamente se distinguen dos perspectivas en la justicia: una ambiental, que se fundamente en los derechos humanos a un ambiente sano y a una mejor calidad de vida, y otra ecológica para los derechos que le corresponden a la Naturaleza. Se repasan sus implicancias en distintas redefiniciones de una comunidad de la justicia ampliada a los seres vivos no-humanos. Se advierte que estos son distintos ensayos en romper el cerco del antropocentrismo característico de la Modernidad.

Evaluation of simulated-LOCA tests that produced large fuel cladding ballooning

International Nuclear Information System (INIS)

Powers, D.A.; Meyer, R.O.

1979-02-01

A description is given of the NRC review and evaluation of simulated-LOCA tests that produced large axially extended ballooing in Zircaloy fuel cladding. Technical summaries are presented on the likelihood of the transient that was used in the tests, the effects of temperature variations on strain localization, and the results of other similar experiments. It is concluded that (a) the large axially extended deformations were an artifact of the experimental technique, (b) current NRC licensing positions are not invalidated by this new information, and (c) no new research programs are needed to study this phenomenon

Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

International Nuclear Information System (INIS)

Sdouz, Gert

2006-01-01

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

Simulation of LOCA and ageing effect with containment liner mockup for analysis of liner-concrete interaction

International Nuclear Information System (INIS)

Wienand, B.; Fila, A.; Hermann, N.; Mueller, M.

2015-01-01

The investigation of the pre-stressed concrete wall behavior including the liner during LOCA conditions is important for the assessment of the structural integrity of the structure and the leak tightness of the liner. In the frame of the NUGENIA ACCEPPT project WP1 G4 'Structural interaction of liner with the concrete', a load test on a reactor containment liner mockup was carried out. The pre-stressed mockup represents a cylindrical part of the liner, embedded in the concrete wall, but without the wall curvature which is not test relevant. It correlates in material and geometrical properties to the EPR containment. The purpose of the test was to check the liners structural behavior and its integrity for Loss of Coolant Accident (LOCA) load combination considering pre-stressing forces and ageing effects due to creep and shrinkage including liner buckling. The test was carried out at the Karlsruhe Institute of Technology (KIT) in September 2013. This article presents the measurement technology, the results and the development of a calculation method for the embedded liner structure. It appears that the liner deformation results are exemplarily shown at the locations of the imperfections, where the liner buckling is anticipated. The measured liner surface strains ranged between +2 and -10 per thousand. The compressive strains are higher than the tensile strains due to the compressive membrane strains caused by pre-stressing and heating. Although the liner got plastic deformations, the liner strains are still far below the elongation at rupture, which indicates that the liner integrity is ensured. We can conclude that the liner mockup test proceeded as planned. The evaluation results show that the purpose of the liner mockup to simulate LOCA + ageing conditions and liner buckling has fully been achieved

Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde; Simulacion de escenarios de accidente severo tipo perdida de refrigerante (Loca) pequeno, con el codigo MELCOR version 2.1 para la central nucleo-electrica de Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft{sup 2} (4.6 cm{sup 2}), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

Naturaleza y cultura en la poesía del modernismo latinoamericano

Directory of Open Access Journals (Sweden)

Rodrigo Javier Caresani

2016-02-01

Full Text Available Desde una constelación de poemas clave —“Caupolicán” (1888 de Rubén Darío, “Amor de ciudad grande” (1882 de José Martí y “En el campo” (1893 de Julián del Casal—, el análisis traza las diferencias en la articulación de una categoría fundante del siglo XIX latinoamericano en lo que va del romanticismo al modernismo. Si el concepto de “civilización” es el soporte del orden emergente que define roles sociales y nacionales en la nueva distribución global del XIX, su carácter intrínsecamente antinómico produce un racimo de oposiciones binarias —igualmente jerárquicas y valorativas— como cultura-naturaleza y ciudad-campo. Bajo estas condiciones, el romanticismo latinoamericano encontró en la nueva relación de la subjetividad con el entorno un camino para ensayar su estética, al re-culturizar el espacio natural para integrarlo al proceso de construcción de identidades nacionales. Desde las coordenadas de una fractura en esta tradición, los poemas del modernismo reconfiguran la antinomia entre naturaleza y cultura en una nueva inflexión que termina afectando los resortes de legitimación de un proyecto estético. En particular, nuestra hipótesis se aproxima al ritmo poético entendiéndolo como una fisura en la “ciudad letrada”, el elemento propulsor de una nueva institución literaria que —mientras transforma el paradigma opositivo civilización-barbarie, original-copia, local-universal— contribuye a deslindar la autoridad del poeta modernista respecto del modelo del letrado tradicional.

Extraction optimization of soluble compounds of yerba maté

Directory of Open Access Journals (Sweden)

César Sambiassi

2002-06-01

Full Text Available The objective of this research was the extraction optimization of water soluble compounds of yerba maté. Measures variables were extract concentration and weight of leaves and twigs. Controlled variables were time and temperature of extraction and water/solid relation. A surface response method of three variables was used as experimental design, with 20 experiences in each case. The range of each variable, defined in the experimental design, was: extraction time, 13.2 to 46.8 minutes; temperature, 48.2 to 81.8°C and water solid relation, 4.64 to 11.36 g water/100 g of dry solid. Extract weight varied from 13.14 to 29.56 g in leaves and 8.98 to 16.32 g in twigs (each one per 100 g of dry solid. Extract concentration varied between 2.17 and 3.43 g/100 ml in leaves and between 1.32 and 2.31 g/100 ml in twigs. The results were fit to a linear equation in each case.O objetivo desta pesquisa foi a otimização da extração aquosa da erva mate. As variáveis medidas foram a concentração do extrato e a massa das folhas e ramos. As variáveis controladas foram: o tempo e a temperatura de extração e a relação água/sólido. Como desenho experimental foi utilizado o método de resposta superficial de três variáveis, com vinte experiências em cada caso. A faixa de variação definida para as variáveis no desenho experimental foram: tempo de extração, de 13,2 a 46,8 minutos; temperatura, de 48,2 a 81,8 °C; relação água/sólido, de 4,64 a 11,36 gramas de água por gramas de sólido seco. A massa do extrato variou de 13,14 a 29,56 gramas para as folhas desramificadas e de 8,98 a 16,32 para os ramos (cada por 100 gramas de sólido seco. A concentração de extrato variou entre 2,17 a 3,43 g/ 100 ml nas folhas e entre 1,32 e 2,31 g/ 100 ml nos ramos. Os resultados foram, em ambos casos, ajustados para uma equação linear.

RELAP 5 Simulations of a hypothetical LOCA in Ringhals 2

International Nuclear Information System (INIS)

Caraher, D.

1987-01-01

RELAP5 simulations of a hypothetical LOCA in Ringhals 2 were conducted in order to determine the sensitivity of the calculated peak cladding temperature (PCT) to Appendix K requirements. The PCT was most sensitive to the assumed model decay heat: Changing from the 1979 ANS Standard to 1.2 times the 1973 Standard increased the PCT by 70 to 100K. After decay heat, the two parameters which affected the PCT the most were steam generator heat transfer and heat transfer lockout. The PCT was not sensitive to the assumed pump rotor condition (locked vs coasting); nor was it sensitive to a modest amount (5 to 10%) of steam generator tube plugging. (author)

The phenomenology of a small break LOCA in a complex thermal hydraulic loop

International Nuclear Information System (INIS)

Di Marzo, M.; Almenas, K.K.; Hsu, Y.Y.; Wang, Z.

1988-01-01

A phenomenological description of the thermal hydraulics events that take place during a simulated Small Break Loss of Coolant Accident (SB-LOCA) is presented. The SB-LOCA transient is described in detail and the various mass and energy transport modes are identified. Similar behavior is observed in other facilities designed for the simulation of this type of accidents. Previous investigations suggest a simple modelling of the phenomena based on fluid mechanic considerations. An extensive experimental program conducted at the experimental facility of the University of Maryland reveals that condensation is a dominant driving force for this type of transients. This finding has significant implications in the modelling of enthalpy transport for some of the flow modes which occur during the transient. In particular it affects the Interruption and Resumption Mode (IRM) during which enthalpy is transported by periodic flow of a two phase mixture. The efforts to predict the flow interruption based on fluid mechanic criteria of phase separation in the hot leg are shown to be misdirected since thermodynamic phenomena taking place in the horizontal portion of the cold legs and in the reactor vessel downcomer are mostly responsible for that transition. For flow resumption to occur the liquid-vapor mixture swelling in the vertical portion of the hot leg determines the occurrence of the liquid spill over the top of the candy cane. (orig.)

Prototype application of best estimate and uncertainty safety analysis methodology to large LOCA analysis

International Nuclear Information System (INIS)

Luxat, J.C.; Huget, R.G.

2001-01-01

Development of a methodology to perform best estimate and uncertainty nuclear safety analysis has been underway at Ontario Power Generation for the past two and one half years. A key driver for the methodology development, and one of the major challenges faced, is the need to re-establish demonstrated safety margins that have progressively been undermined through excessive and compounding conservatism in deterministic analyses. The major focus of the prototyping applications was to quantify the safety margins that exist at the probable range of high power operating conditions, rather than the highly improbable operating states associated with Limit of the Envelope (LOE) assumptions. In LOE, all parameters of significance to the consequences of a postulated accident are assumed to simultaneously deviate to their limiting values. Another equally important objective of the prototyping was to demonstrate the feasibility of conducting safety analysis as an incremental analysis activity, as opposed to a major re-analysis activity. The prototype analysis solely employed prior analyses of Bruce B large break LOCA events - no new computer simulations were undertaken. This is a significant and novel feature of the prototyping work. This methodology framework has been applied to a postulated large break LOCA in a Bruce generating unit on a prototype basis. This paper presents results of the application. (author)

Chlorophyll stability in yerba maté leaves in controlled atmospheres

Directory of Open Access Journals (Sweden)

Rubén O. Morawicki

1999-01-01

Full Text Available The objective of this research was to investigate the stability of chlorophyll in yerba maté leaves in controlled atmospheres of CO2/air mixtures and different water activities at 25°C.Two levels of water activity were selected corresponding to saturated salt solutions of LiCl (a w=0.113 and MgCl2(a w=0.330 and three levels of CO2/air mixtures (0/100,20/80 and 40/60. The chlorophyll content was evaluated using a liquid chromatography HPLC technique. Experimental values varied between 2.16 and 0.61 mg/g of dry matter. For each sample, 5 determination were made during 58 days. Experimental values were fitted to an equation describing a first order reaction. In all cases, the agreement was good with PO objetivo deste trabalho foi pesquisar a estabilidade da clorofila em folhas de erva mate em misturas atmosféricas controladas de CO2/ar e diferentes atividades de vapor de água a 25ºC. Dois níveis de atividade de vapor de água foram selecionadas, correspondendo a soluçoes saturadas de LiCl (a w=0.113 e MgCl2 (a w=0.330 e três níveis de misturas CO2/ar (0/100,20/80 e 40/60. O conteúdo de clorofila foi avaliado usando a técnica de cromatografia líqüida HPLC. Os valores experimentais variaram entre 2.16 e 0.61 mg/g de matéria seca. Para cada amostra foram realizadas 5 determinaçoes durante 58 dias. Os valores experimentais foram ajustados para uma eqüação descrevendo uma reação de primeiro ordem. Em todos os casos houve boa concordância P < 3 10-3. A concentração inicial de clorofila ficou reduzida em média um 30.5% depois de 58 dias. Porém, depois da comparação das constantes de velocidade, não foram achadas diferenças entre elas.

Trastorno por estrés postraumático : naturaleza, evaluación y tratamiento (2009)

OpenAIRE

Bados López, Arturo

2009-01-01

Se abordan diversos aspectos del trastorno por estrés postraumático: naturaleza, edad de comienzo y curso, frecuencia, problemas asociados, génesis y mantenimiento, métodos e instrumentos de evaluación, y eficacia y utilidad clínica del tratamiento psicológico y farmacológico. Además, se ofrecen guías para aplicar los tratamientos psicológicos más eficaces.

Simulating a partial LOCA in a narrow channel using the DSNP simulating system

International Nuclear Information System (INIS)

Saphier, D.

2007-01-01

A partial LOCA accident in a pool type research reactor was investigated. A new MTR type fuel channel model for the DSNP simulation system was developed; permitting detailed axial and radial temperature distribution. New and older heat transfer correlations were incorporated in the model. Simulation for water levels of 14 and 35 cm in a 62 cm channel were performed. The resulting maximum temperatures remain significantly below the aluminium melting point, and no damage to the core will take place under these conditions

Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA

International Nuclear Information System (INIS)

Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.

1983-01-01

Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)

Hydrogen radiolytic production in light and heavy water mixtures under conditions similar to LOCA (loss of coolant accidents)

International Nuclear Information System (INIS)

Garcia Rodenas, L.; Ali, S.P.; Liberman, S.J.

1987-01-01

H 2 , HD and D 2 radiolytic yield in heavy and light water mixtures has been determined to supply the necessary data which will allow to make a realistic estimation of the solution of such gas under LOCA conditions as a function of time. (Author)

Hydrodynamics of AHWR gravity driven water pool under simulated LOCA conditions

International Nuclear Information System (INIS)

Thangamani, I.; Verma, Vishnu; Ali, Seik Mansoor

2015-01-01

The Advanced Heavy Water Reactor (AHWR) employs a double containment concept with a large inventory of water within the Gravity Driven Water Pool (GDWP) located at a high elevation within the primary containment building. GDWP performs several important safety functions in a passive manner, and hence it is essential to understand the hydrodynamics that this pool will be subjected to in case of an accident such as LOCA. In this paper, a detailed thermal hydraulic analysis for AHWR containment transients is presented for postulated LOCA scenarios involving RIH break sizes ranging from 2% to 50%. The analysis is carried out using in-house containment thermal hydraulics code 'CONTRAN'. The blowdown mass and energy discharge data for each break size, along with the geometrical details of the AHWR containment forms the main input for the analysis. Apart from obtaining the pressure and temperature transients within the containment building, the focus of this work is on simulating the hydrodynamic phenomena of vent clearing and pool swell occurring in the GDWP. The variation of several key parameters such as primary containment V1 and V2 volume pressure, temperature and V1-V2 differential pressure with time, BOP rupture time, vent clearing velocity, effect of pool swell on the V2 air-space pressure, GDWP water level etc. are discussed in detail and important findings are highlighted. Further, the effect of neglecting the pool swell phenomenon on the containment transients is also clearly brought out by a comparative study. The numerical studies presented in this paper give insight into containment transients that would be useful to both the system designer as well as the regulator. (author)

Crítica al concepto de naturaleza en el pensamiento sistémico actual: desde las teorías de sistemas no metafísicos

OpenAIRE

Granados Salgado, Diana Marcela

2011-01-01

Tesis (Maestría en Desarrollo Sostenible y Medio Ambiente). Universidad de Manizales. Facultad de Ciencias Contables, Económicas y Administrativas, 2011 Entender el sentido construido y comunicado por los sistemas sociales sobre naturaleza y medio ambiente a través de las diferentes épocas hasta la actualidad, para desde un nuevo horizonte de sentido ganado con la hermenéutica de la historia argumentar un medio ambiente sin naturaleza. Describir las posturas metafísicas o no metafísicas de...

Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde

International Nuclear Information System (INIS)

Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

2013-10-01

In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

Demanda turística internacional por el turismo naturaleza en Costa Rica: Indicadores socio-demográficos y de condición de viaje

Directory of Open Access Journals (Sweden)

Daniel Villalobos-Céspedes

2009-07-01

Full Text Available En Costa Rica, la demanda por actividades turísticas exhibe un posicionamiento relativo del turismo naturaleza. La presente investigación analiza la intención de demanda del turista internacional por turismo naturaleza, según variables sociodemográficas y de condición de viaje. El estudio comprende las modalidades caminata por senderos, visita a volcanes, observación de flora y fauna, observación de aves y canopy. Se sustenta en la base de datos del Instituto Costarricense de Turismo (ICT4, de la Encuesta Aérea del primer trimestre del año 2007 a turistas no residentes mayores de 18 años, y que salieron por el aeropuerto Juan Santamaría. Se encontró que las condiciones viajar con familia, en pareja y procedencia del turista, tienen potencial influencia favorable en la intención de demanda en la mayor parte de esas actividades. Los resultados permiten esbozar una estrategia que orienta las políticas y acciones en materia de turismo naturaleza en el país.

Hydrochemical simulation of a mountain basin under hydrological variability

Science.gov (United States)

Montserrat, S.; Trewhela, T. A.; Navarro, L.; Navarrete, A.; Lagos Zuniga, M. A.; Garcia, A.; Caraballo, M.; Niño, Y.; McPhee, J. P.

2016-12-01

Water quality and the comprehension of hydrochemical phenomena in natural basins should be of complete relevance under hydrological uncertainties. The importance of identifying the main variables that are controlling a natural system and finding a way to predict their behavior under variable scenarios is mandatory to preserve these natural basins. This work presents an interdisciplinary model for the Yerba Loca watershed, a natural reserve basin in the Chilean central Andes. Based on different data sets, provided by public and private campaigns, a natural hydrochemical regime was identified. Yerba Loca is a natural reserve, characterized by the presence of several glaciers and wide sediment deposits crossed by a small low-slope creek in the upper part of the basin that leads to a high-slope narrow channel with less sediment depositions. Most relevant is the geological context around the glaciers, considering that most of them cover hydrothermal zones rich in both sulfides and sulfates, a situation commonly found in the Andes due to volcanic activity. Low pH (around 3), calcium-sulfate water with high concentrations of Iron, Copper and Zinc are found in the upper part of the basin in summer. These values can be attributed to the glaciers melting down and draining of the mentioned country rocks, which provide most of the creek flow in the upper basin. The latter clearly contrasts with the creek outlet, located 18 km downstream, showing near to neutral pH values and lower concentrations of the elements already mentioned. The scope of the present research is to account for the sources of the different hydrological inlets (e.g., rainfall, snow and/or glacier melting) that, depending on their location, may interact with a variety of reactive minerals and generate acid rock drainage (ARD). The inlet water is modeled along the creek using the softwares HEC-RAS and PHREEQC coupled, in order to characterize the water quality and to detect preferred sedimentation sections

Investigation of bubble-condenser operation under large break LOCA conditions

International Nuclear Information System (INIS)

Blinkov, V.; Melikhov, O.; Melikhov, V.; Davydov, M.; Sokolin, A.; Hoffmann, D.; Simon, U.; Bajsz, J.

2000-01-01

In the framework of the PHARE/TACIS project, the experimental test facility for bubble condenser experimental qualification was built at Electrogorsk Research and Engineering Centre. The test facility contains high pressure system, compartments upstream of the bubble condenser and a section of the bubble condenser system. The scaling of the test facility is 1:100. The high pressure system consists of five vessels to appropriately model the leak functions (mass flow rate and enthalpy) during the loss of coolant accidents postulated in the design of VVER-440/V213. Design basis accident (LB LOCA) was experimentally and analytically considered. Results of pre-test analysis with ATHLET and DRASYS codes for determination of necessary test parameters and post-test analysis of three tests are presented. (author)

CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

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Luben Sabotinov

2010-01-01

Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

Large Break LOCA Analysis with New downcomer Nodalizaion and Multi-Dimensional Model and Effect of Cross flow option in MARS code

Energy Technology Data Exchange (ETDEWEB)

Jang, Hyung-wook; Lee, Sang-yong; Oh, Seung-jong; Kim, Woong-bae [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

The phenomena of LOCA have been investigated for long time. The most extensive research project for LOCA was the 2D/3D program experiments. The results of the 2D/3D experiments show flow conditions in the downcomer during end-of-blowdown were highly multi-dimensional at full-scale. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. An LBLOCA analysis for APR1400 with new downcomer input deck was conducted using KREM with MARS-KS 1.4 Version code. Analysis was processed under LBCOCA of 100% break size of cold leg case. The authors developed input deck with new downcomer nodalizaion and Multi-Dimensional downcomer model, then implemented LOCA analysis with new input decks and compared with existing analysis results. PCT from new input and multi-dimensional input deck shows similar PCT trend from original input deck. There occurred more rapid drop of PCT from new and multidimensional input deck than original input deck. PCT from new and multidimensional input deck are satisfied with PCT design limit. It can be concluded that there occurs no acceptance criteria issue even though new and multidimensional input deck are applied to LBLOCA analysis. In future study, comparative analysis with experiment results will be implemented.

Experimental studies on mitigation of LOCA for a high flux research reactor

International Nuclear Information System (INIS)

Saxena, A.K.

2006-01-01

Experimental studies on the rewetting behaviour of hot vertical annular channels were performed to study the mitigation of consequences of loss of coolant accident (LOCA) for a high flux research reactor. Studies were carried out to study the rewetting behaviour with hot inner tube, for bottom flooding and top flow rewetting conditions. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 litres per minute (l pm) (11.7 x 10 -5 m 3 s -1 ). The initial surface temperature of the inner tube was varied from 200 to 500 degC. (author)

Modelling of WWER fuel rod during LOCA conditions using FEM code ANSYS

International Nuclear Information System (INIS)

Bogatyr, S. M.; Krupkin, A. V.; Kuznetsov, V. I.; Novikov, V. V.; Petrov, O. M.; Shestopalov, A. A.

2013-01-01

The report presents the results of the computer simulation of the IFA-650.6 experiment, the sixth test in Halden LOCA test project series, performed in May 18, 2007 with a pre-irradiated WWER-440 fuel with maximum burnup of 56 MWd/kgU. The thermo-mechanical analysis was fulfilled with the license finite element ANSYS code package.The calculation was carried out with the 2D axisymmetric and 3D problem definitions. Analysis of the calculational results shows that the ANSYS code can adequately simulate thermo-mechanical behavior of cladding under IFA-650.6 test conditions. (authors)

Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of exceedance of damage by integrated Safety Analysis Methodology; Arboles de sucesos dinamicos aplicados a secuencias Full Spectrum LOCA. Calculo de la frequencia de excedencia del dano mediante la metodologia Analisis Integrados de Seguridad (ISA)

Energy Technology Data Exchange (ETDEWEB)

Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

2013-09-01

The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Exceedance Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

Analysis and development of the automated emergency algorithm to control primary to secondary LOCA for SUNPP safety upgrading

International Nuclear Information System (INIS)

Kim, V.; Kuznetsov, V.; Balakan, G.; Gromov, G.; Krushynsky, A.; Sholomitsky, S.; Lola, I.

2007-01-01

The paper presents the results of the study conducted to support planned modernization of the South Ukraine nuclear power plant. The objective of the analysis has been to develop the automated emergency control algorithm for primary to secondary LOCA accident for SUNPP WWER-1000 safety upgrading. According to the analyses performed in the framework of safety assesment report, given accident is the most complex for control and has the largest contribution into the core damage frequency value. This is because of initial event diagnostics is difficult, emergency control is complicated for personnel, time available for decision making and actions performing is limited with coolant inventory for make-up, probability of steam dump valves on affected steam generator non-closing after opening is high, and as a consequence containment bypass, irretrievable loss of coolant and radioactive materials release into the environment are possible. Unit design modifications are directed on expansion of safety systems capabilities to overcome given accident and to facilitate the personnel actions on emergency control. Safety systems modification according to developed algorithm will allow to simplify accident control by personnel and enable to control the ECCS discharge limiting pressure below the affected steam generator steam dump valve opening pressure, and decrease the probability of the containment bypass sequences. The analysis of the primary-to-secondary LOCA thermal-hydraulics has been conducted with RELAP5/Mod 3.2, and involved development of the dedicated analytical model, calculations of various plant response accident scenarios, conducting of plant personnel intervention analyses using full-scale simulator, development and justification of the emergency control algorithm aimed on the minimization of negative consequences of the primary-to-secondary LOCA (Authors)

An investigation into fuel pulverization with specific reference to high burn-up LOCA

International Nuclear Information System (INIS)

Yagnik, Suresh; Turnbull, James; Noirot, Jean; Walker, Clive; Hallstadius, Lars; Waeckel, N.; Blanpain, P.

2014-01-01

To investigate the phenomenon of high burn-up fuel pellet material potentially disintegrating into powder under a rapid temperature transient, such as in a LOCA-type accident scenario, two independent scoping studies were commissioned. The first was to investigate the effect of hydrostatic restraint pressure on Fission Gas Release (FGR) from small samples of highly irradiated fuel (71 MWd/kgU) during a series of rapid temperature ramps. Experimentally, when the FGR increased rapidly during the temperature transients, the fuel was assumed to be 'pulverized', i.e., fragmented into powder. In the second series of experiments, laser heating of small samples was used to investigate the temperature at which fuel pulverization was initiated. Subsequent to fuel disintegration, there was always a spectrum of particle sizes present. The significance of this observation was recognized in the context of extended burn-up operation in commercial reactors. Based on the observation from these investigations, a fuel fragmentation threshold has been discussed and developed. We conclude that fuel disintegration could be of potential importance in limiting the performance and productive lifetime of nuclear fuel. However, since only fuel closely adjacent to ballooned or ruptured cladding would be released in a LOCA-type transient, expulsion of pulverized fuel from the ruptured fuel rod is not considered a safety issue; cooling of the defected assembly remains possible and there is no issue with respect to local criticality. (author)

Naturaleza jurídica del encargo fiduciario en la legislación ecuatoriana

OpenAIRE

Irigoyen Samaniego, Daniela

2010-01-01

El presente estudio denominado “Naturaleza Jurídica del Encargo Fiduciario en la Legislación Ecuatoriana” recoge en primer lugar una breve reseña del origen de la figura, tanto a nivel mundial como en los países latinoamericanos y por supuesto de manera especial en Ecuador. Una vez señalados los antecedentes históricos de la figura y su evolución en el ordenamiento jurídico, haremos un análisis de sus características y regulación en la Ley de Mercado de Valores, lo cual nos ...

Investigation on the Behavior of the Jointing Clamps to the Simulated Environmental - LOCA

International Nuclear Information System (INIS)

Ivan, P.; Segarceanu, D.; Geambasu, C.

2002-01-01

The paper presents the main aspect concerning the electric parameter variation of jointing clamps operating under specific environmental conditions determinate by pressure, temperature and humidity. The testing of jointing clamps capability to meet and exceed the required performances all along its operating life implies the performing of LOCA simulation conditions while the jointing clamps is bright in a relatively short time under conditions equivalent to those at the end of its service life. The paper describes ageing and measurement techniques and the analyses of electric parameter behaviour in these environmental simulated conditions. (author)

Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of excedeence of damage by integrated Safety Analysis Methodology

International Nuclear Information System (INIS)

Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

2013-01-01

The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Excedence Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

International Nuclear Information System (INIS)

Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

1981-07-01

The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times

Computation programs for the thermofluidodynamic transient analysis in the containment system following a LOCA

International Nuclear Information System (INIS)

Gorlandi, A.; Mazzini, M.; Oriolo, F.

1979-01-01

This works briefly describes the features of the computation codes available at the Istituto di Impianti Nucleari of the Pisa University for the analysis of the thermofluidodynamic transient in the containment system of a nuclear power plant following a LOCA (RELAP 4/MOD.S, COMPARE, FUMO and CONTEMPT-LT/026). More details are contained in the Annex. Particular attention has been devoted to the opportunity to study, through the computation codes, the effects of the sub division of a full pressure containment system

Thermal stresses at nozzles of nuclear steel containments under LOCA-conditions

International Nuclear Information System (INIS)

Sanchez Sarmiento, G.; Bergmann, A.N.

1986-01-01

During a loss of coolant accident (LOCA) of a PWR-nuclear power plant, a considerable heating of the containment atmosphere is expected to occur. Transient thermal stresses will appear at the containment as a consequence of a non-uniform rise of its temperature. Applying computer codes based on the finite element method, dimensionless general thermal stresses at nozzles of spherical steel containment have been calculated, varying the principal geometrical parameters and the Biot number for the containment internal surface. Atmosphere temperature and Biot number are assumed constant after the accident. Several plots of the maximum principal stresses are provided, which constitute general results applicable to stress analysis of any particular containment of this kind. (orig.)

Naturaleza y alcance del capital social en las comunidades rurales

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Mireya Valdez David

2015-10-01

Full Text Available La presente investigación tuvo como objetivo explicitar la naturaleza y alcance del capital social presente en las relaciones de los integrantes de las comunidades rurales del municipio Andrés Eloy Blanco del Estado Lara (Venezuela. La metodología fue de tipo cualitativa y se utilizaron las técnicas de la Observación Directa, propias de todo proceso de investigación, así como la Observación Participativa, donde el investigador expuso sus puntos de vista con respecto a la información que se recabó. También se utilizó el método Etnográfi co, que aplicado al estilo de vida de un grupo obstaculiza el establecimiento de redes sociales con mayor alcance. Se concluyó entonces que la naturaleza del capital social es espaciada, egocéntrica e informal, propia de los lazos fuertes de las relaciones afectivas; el alcance está caracterizado por relaciones de vinculación y asociación entre comunidades. AbstractThis research aimed to explain the nature and extent of social capital present in relations among the members of a rural communities in Andres Eloy Blanco municipality of Estado de Lara (Venezuela. The methodology was qualitative and used direct observation, which is inherent in all research processes, as well as participatory observation, where the researcher shared his points of view based on the information that was gathered. Ethnographic method was also used. When applied to the lifestyle of a group of people, it hinders the establishment of social networks with the results allowed to make explicit that the nature and of social capital extent in these communities is emerging, which hinders the establishment of social networks. It was concluded that the nature of social capital is spaced, egocentric and informal, own of the ties of affective relationships; the range is characterized by relationships of linkage and association between communities.

Best estimate small break LOCA analysis for KNGR SIS optimization

International Nuclear Information System (INIS)

Song, JIn Ho; Lim, Hong Sik; Bae, Kyoo Hwan; Lee, Joon

1996-01-01

The KNGR has an advanced ECCS design feature which employs four mechanically-separated SI trains where each train consisting of one HPSI pump and one SIT injects ECC water directly into the reactor vessel downcomer annulus. To demonstrate that the KNGR ECCS design features meet the EPRI ALWR requirements of no core uncovery for a break of up to 6 inch diameter, small break LOCA cases with various break sizes were analyzed using a best-estimate analytical procedure. Two kinds of break locations are considered: cold leg and DVI line breaks. It was observed that the KNGR ECCS design can tolerate a cold leg break of up to 10 inches with no core uncovery. However, since DVI line break with 6 inch diameter undergoes slight core uncovery, further investigation is required for KNGR SIS optimization

Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

Energy Technology Data Exchange (ETDEWEB)

Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

2017-05-15

Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

Conservación de la naturaleza en propiedad privada: las Reservas Naturales de la Sociedad Civil en el Valle del Cauca

Directory of Open Access Journals (Sweden)

Melissa Quintero López

2016-01-01

Full Text Available Las Reservas Naturales de la Sociedad Civil en Colombia cumplen un papel importante en la conservación de ecosistemas. En este trabajo se investiga por qué los dueños de áreas privadas en el departamento del Valle del Cauca conservan la naturaleza en sus propiedades. Se consideran las explicaciones entre la teoría económica de la elección racional, la teoría económica sobre los comportamientos altruistas y recíprocos y la teoría de la psicología social sobre los valores de orientación cultural. Se encuentra que las razones predominantes para la conformación de las reservas naturales son el interés propio y la valoración de la naturaleza por parte de sus dueños.

A thermohydraulic analysis for LOCA accident of a CANDU 600 reactor core charged with SEU 43 fuel by means of FIREBIRD code

International Nuclear Information System (INIS)

Serbanel, M.; Catana, A.

2001-01-01

This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters

QUENCH-LOCA program at KIT and results of the QUENCH-L0 bundle test

International Nuclear Information System (INIS)

Stuckert, J.; Grosse, M.; Roessger, C.; Steinbrueck, M.; Walter, M.

2012-01-01

The current LOCA criteria and their safety goals are applied worldwide with minor modifications since the USNRC release in 1973. The criteria are given as limits on peak cladding temperature (T PCT ≤ 1200 C) and on oxidation level ECR (equivalent cladding reacted) calculated as a percentage of cladding oxidized (ECR ≤ 17% calculated using Baker-Just oxidation correlation). These two rules constitute the criterion of cladding embrittlement due to oxygen uptake. The results elaborated worldwide in the 1980s and 1990s on Zircaloy-4 (Zry-4) cladding tubes behavior (oxidation, deformation and bundle coolability) under LOCA conditions constitute a detailed data base and an important input for the safety assessment of LWRs. In-pile test data (with burn-up up to 35 MWd/kgU) were consistent with the out-of-pile data and did not indicate an influence of the nuclear environment on cladding deformation. At high burn-up, fuel rods fabricated from conventional Zry-4 often exhibit significant oxidation, hydriding, and oxide spallation. Thus, many fuel vendors have proposed the use of recently developed cladding alloys, such as M5 registered , ZIRLO trademark and other. Therefore, it is important to verify the safety margins for high burn-up fuel and fuel claddings with new alloys. Due to long cladding hydriding period for the high fuel burn-up, post-quench ductility is strongly influenced not only by oxidation but also hydrogen uptake. The 17% ECR limit is inadequate to ensure post-quench ductility at hydrogen concentrations higher than ∼500 wppm. Due to so called secondary hydriding (during oxidation of inner cladding surface after burst), which was firstly observed in JAEA, the hydrogen content can reach 4000 wppm in Zircaloy cladding regions around burst. To investigate the influence of these phenomena on the applicability of the embrittlement criteria for the German nuclear reactors it was decided to perform the QUENCH-LOCA bundle test series at the Karlsruhe Institute

QUENCH-LOCA program at KIT and results of the QUENCH-L0 bundle test

Energy Technology Data Exchange (ETDEWEB)

Stuckert, J.; Grosse, M.; Roessger, C.; Steinbrueck, M.; Walter, M. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

2012-11-01

The current LOCA criteria and their safety goals are applied worldwide with minor modifications since the USNRC release in 1973. The criteria are given as limits on peak cladding temperature (T{sub PCT} {<=} 1200 C) and on oxidation level ECR (equivalent cladding reacted) calculated as a percentage of cladding oxidized (ECR {<=} 17% calculated using Baker-Just oxidation correlation). These two rules constitute the criterion of cladding embrittlement due to oxygen uptake. The results elaborated worldwide in the 1980s and 1990s on Zircaloy-4 (Zry-4) cladding tubes behavior (oxidation, deformation and bundle coolability) under LOCA conditions constitute a detailed data base and an important input for the safety assessment of LWRs. In-pile test data (with burn-up up to 35 MWd/kgU) were consistent with the out-of-pile data and did not indicate an influence of the nuclear environment on cladding deformation. At high burn-up, fuel rods fabricated from conventional Zry-4 often exhibit significant oxidation, hydriding, and oxide spallation. Thus, many fuel vendors have proposed the use of recently developed cladding alloys, such as M5 {sup registered}, ZIRLO trademark and other. Therefore, it is important to verify the safety margins for high burn-up fuel and fuel claddings with new alloys. Due to long cladding hydriding period for the high fuel burn-up, post-quench ductility is strongly influenced not only by oxidation but also hydrogen uptake. The 17% ECR limit is inadequate to ensure post-quench ductility at hydrogen concentrations higher than {approx}500 wppm. Due to so called secondary hydriding (during oxidation of inner cladding surface after burst), which was firstly observed in JAEA, the hydrogen content can reach 4000 wppm in Zircaloy cladding regions around burst. To investigate the influence of these phenomena on the applicability of the embrittlement criteria for the German nuclear reactors it was decided to perform the QUENCH-LOCA bundle test series at the

Trabajando el acercamiento a la naturaleza de los niños y niñas en el Grado de Educación Infantil. Crucial en la sociedad actual

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Jerónimo Torres-Porras

2017-01-01

Full Text Available Los niños y niñas de la sociedad actual se encuentran alejados de entornos naturales y están necesitados de contacto con la naturaleza, lo que puede ser el motivo de distintos síntomas como estrés o ansiedad. Existen investigaciones que muestran que el contacto con la naturaleza es beneficioso para el desarrollo de los infantes además de fomentar una actitud más pro-ambiental. En este trabajo se pretende concienciar al alumnado del Grado de Educación Infantil de dichos beneficios mediante la ejecución de una serie de actividades centradas en el uso de los parques urbanos como recurso para el acercamiento a la naturaleza de los niños y las niñas de la etapa de infantil. Es necesario que tanto universidad, escuela y familia fomenten el uso y disfrute de estos espacios.

Comparison of the DVI line break LOCA with the equivalent cold leg break with the ATLAS facility

International Nuclear Information System (INIS)

Choi, K. Y.; Cho, S.; Kang, K. H.; Park, H. S.; Kim, Y. S.; Baek, W. P.

2010-01-01

The APR1400 (Advanced Power Reactor, 1400 MWe) adopts a DVI (Direct Vessel Injection) method for ECC (Emergency Core Cooling) water delivery rather than a conventional CLI (Cold Leg Injection) method as an advanced safety feature. The break scenario of one DVI nozzle is taken into account in the small break LOCA analysis. Transient behavior during the DVI line breaks needs to be investigated and compared with the equivalent break on the cold leg. An 8.5-inch double-ended break of one DVI nozzle was simulated with the ATLAS, and a counterpart test for the DVI break was performed at the cold leg with the equivalent break size for comparison. This comparison will contribute to enhancing a comprehensive understanding of the thermal hydraulic behavior during transients. A constructed integral effect database is also used to validate the existing conservative safety analysis methodology and to develop a best-estimate safety analysis methodology for small-break LOCAs. A post-test calculation was performed with a best-estimate safety analysis code, MARS 3.1, in order to examine its prediction capability and to identify any code deficiencies for thermal hydraulic phenomena occurring during the transient. (authors)

Development and application of KEPRI realistic evaluation methodology (KREM) for LB-LOCA

International Nuclear Information System (INIS)

Ban, Chang-Hwan; Lee, Sang-Yong; Sung, Chang-Kyung

2004-01-01

A realistic evaluation method for LB-LOCA of a PWR, KREM, is developed and its applicability is confirmed to a 3-loop Westinghouse plant in Korea. The method uses a combined code of CONTEMPT4/MOD5 and a modified RELAP5/MOD3.1. RELAP5 code calculates system thermal hydraulics with the containment backpressure calculated by CONTEMPT4, exchanging the mass/energy release and backpressure in every time step of RELAP5. The method is developed strictly following the philosophy of CSAU with a few improvements and differences. Elements and steps of KREM are shown in Figure this paper. Three elements of CSAU are maintained and the first element has no differences. An additional step of 'Check of Experimental Data Covering (EDC)' is embedded in element 2 in order to confirm the validity of code uncertainty parameters before applying them to plant calculations. The main idea to develop the EDC is to extrapolate the code accuracy which is determined in step 8 to the uncertainties of plant calculations. EDC is described in detail elsewhere and the basic concepts are explained in the later section of this paper. KREM adopts nonparametric statistics to quantify the overall uncertainty of a LB-LOCA at 95% probability and 95% confidence level from 59 plant calculations according to Wilks formula. These 59 calculations are performed in step 12 using code parameters determined in steps 8 and 9 and operation parameters from step 11. Scale biases are also evaluated in this step using the information of step 10. Uncertainties of code models and operation conditions are reflected in 59 plant calculations as multipliers to relevant parameters in the code or as input values simply. This paper gives the explanation on the overall structures of KREM and emphasizes its unique features. In addition, its applicability is confirmed to a 3-loop plant in Korea. KREM is developed for the realistic evaluation of LB-LOCA and its applicability is successfully demonstrated for the 3-loop power plants in

Numerical Ballooning and Burst Prediction of Fuel Cladding During LOCA Transients in LWR

International Nuclear Information System (INIS)

Landau, E.; Weiss, Y.; Szanto, M.

2014-01-01

Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis, most former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. Several codes for reactor core and fuel rod analysis exist nowadays, such as FRAPTRAN1.4 or RELAP5-3D . These are tailor-made codes, designed to predict general core behavior and fuel performance, and while they are also used in predicting core components behavior during accident conditions, including those of cladding ballooning and failure with good accuracy, they contain several limitations on modeling the full transient cladding thermo mechanical phenomena. Limitations such as mechanical models being one dimensional or in axisymmetric geometries only, relying mostly on analytical models therefore having further restricting assumptions in return for accuracy, etc. These limitations disable the simulation of several important aspects, such as modeling 3D azimuthal behavior for example. The objective of the current work is to develop a comprehensive numerical model for predicting zircalloy cladding thermo mechanical behavior during a LOCA. The model will eventually predicts full cladding ballooning and burst behavior followed by fuel relocation, for fuel rods that can be subjected to 3D distributed flux. The model is fully three dimensional and is created using the commercial FEM numerical simulation software ABAQUS© applying

An outcome of nuclear safety research in JAERI. Case study for LOCA, FP, criticality and reprocessing

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Ito, Keishiro; Kawashima, Kei; Katsuki, Chisato; Shirabe, Masashi

2009-09-01

An outcome of nuclear safety research done by JAERI was case studied by the bibliometric method. (1) For LOCA (loss-of-coolant accident) a domestic share of JAERI in monoclinic research paper was 63% at the past (20) 1978-1982 but was decreased to 40% at the present 1998-2002. For co-authored papers a domestic share between JAERI and PS (public sectors) is almost zero at past (20) but increased to 4% at the present. Research cooperation is active between Tokyo University and JAERI or between JAERI and Nagoya University. (2) Project-type research is to have a large monopolization in papers and that of basic-type research is to have a large development of research networking (DRN). (3) For FP, a share of co-authored paper is high due to an enhanced cooperation among JAERI-PO (Public Organization)-PS. For criticality, research activity was enhanced after JCO accident, especially at NUCEF. (4) For reprocessing, PS had a monopolistic position with a domestic share of 71% and a share of JAERI was about 20%. (5) LOCA and RIA outputs born by NSR-JAERI coincided partly to those of the Safety Licensing Guidelines but a share of contribution done by JAERI was ambiguous due to the lack of necessary information. (author)

A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

Directory of Open Access Journals (Sweden)

Martina Adorni

2011-01-01

Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

Reflooding phase of the LOCA - state of the art II. Rewetting and liquid entrainment

International Nuclear Information System (INIS)

Elias, E.; Yadigaroglu, G.

1977-01-01

Understanding the mechanisms by which hot fuel rods quench and the physics of liquid droplet entrainment is important for the analysis of the reflooding phase of the LOCA. Published models of the rewetting process include simple one-dimensional solutions. The basic physical assumptions of these models and the numerical values assigned to the various parameters, as well as empirical rewetting correlations are discussed. The various mechanisms for liquid droplet entrainment and analytical formulations of the critical gas velocity and of the droplet diameter at the onset of entrainment are reviewed

Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

International Nuclear Information System (INIS)

Bozoki, G.; Kohut, P.; Fitzpatrick, R.

1989-02-01

This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

Measurement of mist cooling of PWR during LOCA by LDA

International Nuclear Information System (INIS)

Lee, S.L.; Sheen, H.J.; Issapour, I.

1985-01-01

The prediction of temperature distribution and heat transfer within rod bundles during the refill and reflood phase of a LOCA (loss of coolant accident) is of critical importance for determining the location and size of blockages due to clad deformation in a pressurized water reactor (PWR). Mist cooling by small droplets generated from large droplets on hitting grid spacers has been suggested as one of the most important heat transfer mechanisms which are responsible for the development of this temperature transient. The questions to be asked are whether such small droplets indeed exist and, if so, how are they related to the cooling of the fuel rods. Hereby reported is the result of a direct experimental investigation on these questions by a special laser-Doppler anemometry (LDA) particle sizing technique together with temperature measurements of the rod claddings and flow in the subchannel

Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

International Nuclear Information System (INIS)

Axford, D.J.; Martin, I.C.; McAuley, S.J.

1981-04-01

This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, Thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCE is expected to closely model a LPWR LOCA. 2 - Description of test: Experiment LP-LB-1 was conducted on 3 February 1984 in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Cooperation and Development. The primary objectives of Experiment LP-LB-1 were to determine system transient characteristics and to assess code predictive capabilities for design basis large-break loss-of-coolant accidents in pressurized water reactors (PWRs). This experiment simulated a double-ended offset shear of one inlet pipe in a four-loop PWR and was initiated from conditions representative of licensing limits in a PWR. Other boundary conditions for the simulation were loss of offsite power, rapid primary coolant pump coast down, and United Kingdom minimum safeguard emergency core coolant injection rates. The nuclear fuel rods were not pressurized. The transient was initiated by opening the quick-opening blowdown valves in the broken loop hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

El debate sobre las necesidades, y la cuestión de la “naturaleza humana”

OpenAIRE

Razeto, Luis

2011-01-01

En el artículo se aborda la cuestión filosófica de la “naturaleza humana”, de cuya respuesta depende la posibilidad de cambios económicos, políticos y sociales, más o menos profundos y extendidos. En tal contexto, se parte examinando el comportamiento del consumidor moderno, cuyas necesidades, aspiraciones y deseos y el modo de satisfacerlos actualmente, parecieran constituir un obstáculo insalvable para generar tales procesos de cambio.  Si el modo de ser del consumidor moderno fuese expresi...

Safety Significance of the Halden IFA-650 LOCA Test Results

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Petit, Marc; Hozer, Zoltan; Kelppe, Seppo; Khvostov, Grigori; Hafidi, Biya; Therache, Benjamin; Heins, Lothar; Valach, Mojmir; Voglewede, John; Wiesenack, Wolfgang

2010-01-01

The safety criteria for loss-of-coolant accidents were defined to ensure that the core would remain coolable. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have generated a need to re-examine these criteria and to verify their continued validity. As part of international efforts to this end, the OECD Halden Reactor Project program implemented a LOCA test series. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, IRSN, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The fourth test of the series, IFA-650.4 conducted in April 2006, caused particular attention in the international nuclear community. The fuel used in the experiment had a high burnup, 92 MWd/kgU, and a low pre-test hydrogen content of about 50 ppm. The test aimed at and achieved a peak cladding temperature of 850 deg. C. The rod burst occurred at 790 deg. C. The burst caused a marked temperature increase at the lower end and a decrease at the upper end of the system, indicating that fuel relocation had occurred. Subsequent gamma scanning showed that approximately 19 cm of the fuel stack were missing from the upper part of the rod and that fuel had fallen to the bottom of the capsule. PIE at the IFE-Kjeller hot cells corroborated this evidence of substantial fuel relocation. The fact that fuel dispersal could occur upon ballooning and burst, i.e. at cladding temperatures as low as 800 deg. C and thus far lower than the temperature entailed by the current 1200 deg. C / 17% ECR limit, caused concern. The

Application of realistic (best- estimate) methodologies for large break loss of coolant (LOCA) safety analysis: licensing of Westinghouse ASTRUM evaluation model in Spain

International Nuclear Information System (INIS)

Lage, Carlos; Frepoli, Cesare

2010-01-01

When the LOCA Final Acceptance Criteria for Light Water Reactors was issued in Appendix K of 10CFR50 both the USNRC and the industry recognized that the rule was highly conservative. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the USNRC began a research program to identify the degree of conservatism in those models permitted in the Appendix K rule and to develop improved thermal-hydraulic computer codes so that realistic accident analysis calculations could be performed. The overall results of this research program quantified the conservatism in the Appendix K rule and confirmed that some relaxation of the rule can be made without a loss in safety to the public. Also, from a risk-informed perspective it is recognized that conservatism is not always a complete defense for lack of sophistication in models. In 1988, as a result of the improved understanding of LOCA phenomena, the USNRC staff amended the requirements of 10 CFR 50.46 and Appendix K, 'ECCS Evaluation Models', so that a realistic evaluation model may be used to analyze the performance of the ECCS during a hypothetical LOCA. Under the amended rules, best-estimate plus uncertainty (BEPU) thermal-hydraulic analysis may be used in place of the overly prescriptive set of models mandated by Appendix K rule. Further guidance for the use of best-estimate codes was provided in Regulatory Guide 1.157 To demonstrate use of the revised ECCS rule, the USNRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology as an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis. More recently the CSAU principles have been generalized in the Evaluation Model Development and Assessment Process (EMDAP) of Regulatory Guide 1.203. ASTRUM is the Westinghouse Best Estimate Large Break LOCA evaluation model applicable to two-, three

Fission product release from high gap-inventory LWR fuel under LOCA conditions

International Nuclear Information System (INIS)

Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

1980-01-01

Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

The fuel and channel thermal/mechanical behaviour code FACTAR 2.0 (LOCA)

International Nuclear Information System (INIS)

Westbye, C.J.; Mackinnon, J.C.; Gu, B.W.

1996-01-01

The computer code FACTAR 2.0 (LOCA) models the thermal and mechanical response of components within a single CANDU fuel channel under loss-of-coolant accident conditions. This code version is the successor to the FACTAR 1.x code series, and features many modelling enhancements over its predecessor. In particular, the thermal hydraulic treatment has been extended to model reverse and bi-directional coolant flow, and the axial variation in coolant flow rate. Thermal radiation is calculated by a detailed surface-to-surface model, and the ability to represent a greater range of geometries (including experimental configurations employed in code validation) has been implemented. Details of these new code treatments are described in this paper. (author)

LOCA, LOFA and LOVA analyses pertaining to NET/ITER safety design guidance

International Nuclear Information System (INIS)

Ebert, E.; Raeder, J.

1991-01-01

The analyses presented pertain to loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA). These types of accidents may jeopardise components and plasma vessel integrity and cause radioactivity mobilisation. The analyses reviewed have been performed under the assumption that the plasma facing components are protected by a carbon based armour. Accidental temperatures and pressure transients are quantified, the possibility of reaction products combustion is investigated and worst case accidental public doses are assessed. On this basis, design recommendations are given and design features such as low plasma facing components armour temperatures (on almost the entire surface) and inert gas adjacent to the vacuum vessel have been implemented. (orig.)

Modeling the cool down of the primary heat transport system using shut down cooling system in normal operation and after events such as LOCA

International Nuclear Information System (INIS)

Icleanu, D.L.; Prisecaru, I.

2015-01-01

This paper aims at modeling the cooling of the primary heat transport system using shutdown cooling system (SDCS), for a CANDU 6 NPP in all operating modes, normal and abnormal (particularly in case of LOCA accident), using the Flowmaster calculation code. The modelling of heavy water flow through the shutdown cooling system and primary heat transport system was performed to determine the distribution of flows, pressure in various areas of the hydraulic circuit and the pressure loss corresponding to the components but also for the heat calculation of the heat exchangers related to the system. The results of the thermo-hydraulic analysis show that in all cases analyzed, normal operation and for LOCA accident regime, the performance requirements are confirmed by analysis

El crecimiento del ecoturismo y de las actividades físicas de aventura en la naturaleza (afan: elementos para comprender la situación actual en España y Brasil

Directory of Open Access Journals (Sweden)

Ana Márcia Silva

2008-12-01

Full Text Available Este artículo presenta algunos datos sobre el crecimiento del ecoturismo y en directa relación con la emergente difusión e implantación de las actividades físicas de aventura en la naturaleza (AFAN en los ámbitos territoriales de España y Brasil. Sobre estos datos tratamos de identificar, en la producción académica sobre el tema, los motivos del incremento del interés por estas actividades y sus influencias en el ser humano y en la naturaleza. El ecoturismo es la dimensión del turismo que más crece, con la consolidación de diferentes tipos de actividades físicas de aventura en la naturaleza. En este trabajo, ecoturismo se interpreta como una articulación ética y política entre las dimensiones del medio ambiente, de las relaciones sociales y de la subjetividad humana, desarrollada en espacios naturales e iniciando una nueva relación ser humano/naturaleza, ahora no por la dominación sino por la vía de la interacción, en un proceso de desarrollo endógeno cimentado en la autonomía de las poblaciones involucradas.

Revisiting large break LOCA with the CATHARE-3 three-field model

International Nuclear Information System (INIS)

Valette, Michel; Pouvreau, Jérôme; Bestion, Dominique; Emonot, Philippe

2011-01-01

Highlights: ► CATHARE 3 enables a three-field analysis of a LB LOCA. ► Reflooding experiments in isolated rod bundles are satisfactory predicted. ► A BETHSY integral test simulation supports the CATHARE 3 3-field assessment. - Abstract: Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in a rod bundle is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit: core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic six-equation model is used in the other parts of the loop. An analysis of these first results is presented and future work is defined for improving the droplet behavior simulation in both the upper plenum and the hot legs.

Estimation of Leak Flow Rate during Post-LOCA Using Cascaded Fuzzy Neural Networks

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong Yeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

2016-10-15

In this study, important parameters such as the break position, size, and leak flow rate of loss of coolant accidents (LOCAs), provide operators with essential information for recovering the cooling capability of the nuclear reactor core, for preventing the reactor core from melting down, and for managing severe accidents effectively. Leak flow rate should consist of break size, differential pressure, temperature, and so on (where differential pressure means difference between internal and external reactor vessel pressure). The leak flow rate is strongly dependent on the break size and the differential pressure, but the break size is not measured and the integrity of pressure sensors is not assured in severe circumstances. In this paper, a cascaded fuzzy neural network (CFNN) model is appropriately proposed to estimate the leak flow rate out of break, which has a direct impact on the important times (time approaching the core exit temperature that exceeds 1200 .deg. F, core uncover time, reactor vessel failure time, etc.). The CFNN is a data-based model, it requires data to develop and verify itself. Because few actual severe accident data exist, it is essential to obtain the data required in the proposed model using numerical simulations. In this study, a CFNN model was developed to predict the leak flow rate before proceeding to severe LOCAs. The simulations showed that the developed CFNN model accurately predicted the leak flow rate with less error than 0.5%. The CFNN model is much better than FNN model under the same conditions, such as the same fuzzy rules. At the result of comparison, the RMS errors of the CFNN model were reduced by approximately 82 ~ 97% of those of the FNN model.

ASTEC and MELCOR comparison for a VVER-1000 60 mm small break LOCA

International Nuclear Information System (INIS)

Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

2005-01-01

In this paper a comparison between severe accident calculations performed for a WWER 1000 with the ASTEC1.1v0 and MELCOR 1.8.5 computer codes for a small break LOCA (ID 60 mm) without intervention of hydro accumulators is presented. This investigation has been performed in the framework of the SARNET project under the EURATOM 6th framework program. Once the accident sequence scenario is specified, both codes (MELCORE and ASTEC) are able to determine the core and containment damaged states, to estimate the release of radionuclides from the fuel as well as from the primary circuit and containment. Theses results are used to estimate the maximum period of the time during which the personnel could still take particular decisions in order to mitigate such an accident. The aim of the performed analysis is to estimate the discrepancy between ASTEC and MELCORE 1.8.5 calculations. Such discrepancies will be studied, if the case, proposal for ASTEC improvements will be made. Also the ASTEC capability to simulate specific reactor accident scenarios and/or particular safety systems will be tested. The final target is to propose severe accident management procedure for WWER 1000 reactors. In conclusions, the analysis for a small break LOCA (ID 60 mm without hydroelectricities) has shown some discrepancies between ASTEC and MELCORE especially during the degradation of the core. Further analyses are planed in which the MELCORE temperature 'set point' for core degradation (2520 K) will be progressively increased to approach the ASTEC one (which has been estimated to be about 3200 K). The comparison of the new results will allow a better evaluation of the in-vessel models implemented in ASTEC

Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements

International Nuclear Information System (INIS)

Wilkowski, G.; Rudland, D.; Wolterman, R.; Krishnaswamy, P.; Scott, P.; Rahman, S.; Fairbanks, C.

2002-01-01

The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, A cceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors . These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from the

Estudio hidrogeológico y de calidad de agua en el sector oriental de la Sierra de San Javier entre las localidades de Yerba Buena y el Manantial. Provincia de Tucuman, Republica Argentina

Directory of Open Access Journals (Sweden)

D’Urso, C. H.

2005-12-01

Full Text Available The study area is located in the center west of the county of Tucumán, in the NW of Argentina. It extends from the oriental border of San Javier’s mountain between the towns of Yerba Buena and Manantial. In the piedmont zone and plain, have settled down diverse urban, agricultural centers and an important industrial complex that are supplied by underground water. The objective of this paper is to define the different geologic factors that impact in the behavior of the underground water and to define areas with appropriate hydrogeological characteristics for its use. The investigation determined that the water contained in the aquifers is of good quality and due to the high permeability, important flows can be obtained, that guarantee the supply to the population, agriculture and industry.La zona de estudio está ubicada en el centro oeste de la provincia de Tucumán, en el noroeste de Argentina. Se extiende desde el borde oriental de la sierra de San Javier entre las localidades de Yerba Buena y el Manantial. En el piedemonte y llanura se han establecido diversos centros urbanos, agrícolas y un importante complejo industrial que se abastecen de agua subterránea. El objetivo del trabajo es delimitar los distintos factores geológicos que inciden en el comportamiento del agua subterránea y tratar de definir zonas con características hidrogeológicas apropiadas para su aprovechamiento. De esta investigación se determinó que el agua que contienen los acuíferos es de buena calidad y debido a la elevada permeabilidad de los mismos, se pueden obtener importantes caudales, con lo que se garantiza el abastecimiento a la población, agricultura e industria.

THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

International Nuclear Information System (INIS)

Asahi, Yoshiro

2001-01-01

1 - Description of problem or function: THYDE-P1 analyzes the behaviour of LWR plants in response to various disturbances, including the thermal hydraulic transient following a break of the primary coolant pipe system, generally referred to as a loss-of-coolant-accident (LOCA). LOCA can be considered as the most critical condition for testing the methods and models for plant dynamics, since thermal hydraulic conditions in the system change drastically during the transient. THYDE-P is capable of a complete LOCA calculation from start to complete reflooding of the core by subcooled water. The program performs steady-state adjustment, which is complete in the sense that the steady state obtained is a set of exact solutions of all the transient equations without time derivatives, not only for plant hydraulics but also for all the other phenomena in the simulation of a PWR plant. THYDE-P2 contains among others the following improvements over THYDE-P1: (1) not only the mass and momentum equations but also the energy equation are included in the non-linear implicit scheme; (2) the valve model is implemented; (3) the relaxation equation for void fraction is theoretically derived; (4) vectorized programming is implemented; (5) both EM (evaluation mode) and BE (best estimate) calculations are possible. THYDE-W is an improved version of THYDE-P2 and contains the following additional features: (a) analysis of multiple number of disjoint loops is possible; (b) a control system simulation model is included; (c) the trip model has been improved; (d) heavy water is allowed as coolant; (e) the effect of drift flux is accounted for in the steady state calculation; (f) boron transport is included; (g) to obtain steady state loop heat balance, the option of adjusting the enthalpy distribution is prepared included in addition to that of adjusting heat exchanger areas; (h) to obtain steady state pressure distribution, three other options are prepared in addition to the original ones

LDA measurement of droplet behavior across tie plate during dispersed flow portion of loca reflood

International Nuclear Information System (INIS)

Lee, S.L.; Srinivasan, J.; Cho, S.K.

1980-01-01

The flow of an air-water droplet dispersion in a simulated 3-D test section in the reflood portion of LOCA was studied. For this purpose, a new scheme of Laser-Doppler Anemometry for the simultaneous measurement of size and velocity of large-size [0.5 mm-6 mm] droplets was developed and utilized. It was observed that the size distribution of the reentrained droplets depends mainly on the flow regimes and is essentially independent of that of the incoming dispersion below the tie plate. 8 refs

Naturaleza y dimensión del rezago habitacional en México

Directory of Open Access Journals (Sweden)

Ignacio Kunz-Bolaños

2008-01-01

Full Text Available El presente artículo estima la dimensión y naturaleza del rezago de vivienda en México. La mayoría de los cálculos no identifican plenamente las deficiencias derivadas del hacinamiento y tampoco las carencias que se producen por la mala calidad de las viviendas. En este trabajo se estiman cuatro grandes tipos de déficit: a las unidades que no cumplen con las condiciones mínimas para definirlas como vivienda (2 millones; b el déficit producto de las extensiones familiares y hogares extras (8 millones; c el déficit cualitativo, es decir, el que resulta de la mala condición o escaso equipamiento de las viviendas (10.3 millones; y d otras situaciones no deseadas (0.7 millones. Para realizar la estimación se utilizó la muestra censal del año 2000.

Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients

International Nuclear Information System (INIS)

Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

1980-01-01

The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations

Sociedad y naturaleza en la mitología Muisca

Directory of Open Access Journals (Sweden)

François Correa Rubio

2005-01-01

Full Text Available Analizando la mitología de los muiscas ilustraré cómo se trata de un discurso que puede guiar el comportamiento social porque esta constituido por un conjunto de abstracciones que, en el lenguaje simbólico, expresan el conocimiento de la sociedad sobre sus propias relaciones y de estas con la naturaleza. El análisis evidencia cómo el sol y la luna eran los símbolos dominantes no sólo por su importancia en el ordenamiento del tiempo y espacio del universo y, en consecuencia, en las tareas cotidianas, sino cómo su comportamiento pretendía ser explicado a semejanza de la estabilidad y variabilidad de las relaciones sociales. Argumentaré que esta analogía entre el comportamiento de los astros y las gentes constituye un operador lógico en el que descansa la simbología. Los mitos no son, pues, fantasía, ni en ellos «todo puede suceder». Son construcciones culturales que expresan en un discurso codificado, conocimientos de la experiencia que la sociedad debe seguir para garantizar su propia reproducción social.

RALLY… NATURALEZA, RECREACIÓN Y SALUD

Directory of Open Access Journals (Sweden)

María Morera Castro

2007-07-01

Full Text Available El rally es una actividad que congrega a un grupo de personas. Las cuales conforman equipos, con el propósito de recorrer un territorio en un tiempo determinado, realizando acciones (pistas, acertijos o desafíos que les permitan avanzar hasta lograr el cumplimiento del objetivo planteado. Nadie conoce en dónde están las pistas, acertijos o desafíos. Las reglas del juego se les explican a todos los participantes, al igual que las normas de seguridad. Se debe tomar en cuenta al realizar un rally, el objetivo, el tema que se quiere trabajar, el lugar donde se va a realizar, los materiales que se van a necesitar y con los que se cuenta, también las pistas acertijos o desafíos que se pueden incluir. A su vez, como se supervisará que todos los equipos pasen por los puestos de control. Como cualquier actividad que lleve una pizca de competitividad, el juego limpio y el respeto debe de prevalecer ante cualquier interés. Si por algún motivo, alguno de los equipos tuvo alguna emergencia, la solidaridad debe reinar y la actividad pasará a ser un aprendizaje de valores. El sentido aventurero es parte esencial en esta actividad. El reto por lo desconocido, lo nuevo, mantiene en alto el deseo de retarse a sí mismo y al grupo. Esta es una actividad que da cabida a que grupos de amigos y amigas, niños, niñas, adultos, familias, entre otros, convivan y pasen un rato juntos, un día agradable, sano y en contacto con la naturaleza. Rescatando la cooperación, astucia y sobre todo el trabajo en equipo.

A comparison of LOCA analysis using SMOKIN and CERBERUS codes

Energy Technology Data Exchange (ETDEWEB)

Younis, M H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Gaboury, G [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

This paper presents the results of a comparison of the analyses of a postulated Loss of Coolant Accident (LOCA) in Pickering NGS A reactors using the two neutron kinetics codes SMOKIN and CERBERUS. Both codes have been used to simulate the space-time neutronic behaviour of CANDU-PHWR reactors. The main objective of the present study is to evaluate the accuracy with which SMOKIN can predict power transients compared to CERBERUS. The comparison shows that the two codes produce similar bulk power and reactivity transients. However, SMOKIN was found to overestimate the power transient (relative to CERBERUS) in some regions of the core, which is indicative of the spatial differences between the two codes. It was demonstrated that part of this overestimate is due to the use of reaction-rate averaged fuel properties in SMOKIN, compared to instantaneous fuel properties in CERBERUS. (author). 5 refs., 3 tabs., 6 figs.

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

Energy Technology Data Exchange (ETDEWEB)

Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

International Nuclear Information System (INIS)

Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

2017-01-01

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2

International Nuclear Information System (INIS)

Rychkov, M.; Chikkanagoudar, U.; Sehgal, B.R.

2004-01-01

A RELAP5 model for the analysis of the PSB-VVER test facility was developed by EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, we have modified the PSB-VVER facility's RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5's calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the '11% UP LOCA' test data, the RELAP5/MOD3.2 model was used for a so-called 'blind' transient calculation of the test '2*25% HL LOCA' and the results obtained were compared with the experimental data provided after the calculation. From the results of the qualitative and quantitative comparison of the 2 test calculations and the experimental data, we can state that the RELAP5/MOD3.2 code gives a satisfactory modeling of the PSB-VVER facility' thermal hydraulic phenomena

Experimental results of the effective water head in downcomer during reflood phase of a PWR LOCA

International Nuclear Information System (INIS)

Sudo, Yukio; Murao, Yoshio; Akimoto, Hajime

1980-08-01

The results and analysis of an experiment for the effective water head in downcomer with 50mm gap size are described. The main objective of the experiment was to clarify the effect of gap size on reflooding in a PWR LOCA. The effective water head in downcomer is the driving force for feeding emergency coolant into the core during reflood phase of a PWR LOCA. Discussions presented here follow those of a previous report in which experimental results and analysis were described for the case of 200mm gap size. Experimental Conditions were: Initial Wall Temperature = 200 -- 300 0 C, Back Pressure = 1 atm., Coolant Temperature = 71 -- 100 0 C, Extraction Water Velocity = 0 -- 2 cm/s, Gap Size = 50 mm. The effective water head history obtained in the experiment was compared with those predicted with Sudo's void fraction correlation. In the prediction, heat input to coolant was calculated from the response of measured wall temperature with heat condition analysis. The experimental results and analysis reveals that: (1) The effects of the gap size and initial wall temperature are evident, (2) The effect of extraction water velocity is negligible, and (3) The predicted history of effective water head is in good agreement with the experimental results except during the transient period in which the effective water head is descreasing. (author)

Proyectar con la naturaleza mediante la Metodología de los Estudios de Impacto Ambiental en ordenaciones residenciales

OpenAIRE

Higueras García, Esther

2013-01-01

Proyectar con la naturaleza mediante la Metodología de los Estudios de Impacto Ambiental en ordenaciones residenciales .- Procedimiento de acción en la planificación ambiental .- Las técnicas de agregación de impactos .- Las afecciones de los planes de ordenación sobre el territorio. .- Las medidas preventivas y correctoras de planes .- Evaluación critica de los estudios de impacto ambiental

Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

International Nuclear Information System (INIS)

Lelong, Franck

2010-01-01

In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

La naturaleza jurídica de la posibilidad de prescripción de los delitos

OpenAIRE

Cerrada Moreno, Manuel

2017-01-01

Tradicionalmente, la prescripción de los delitos ha sido estudiada como una institución a la que se ha asignado una naturaleza jurídica material o procesal para derivar de ello consecuencias mediante el sistema lógico-deductivo, pero de este modo se llega a resultados que no son válidos, pues contradicen lo dispuesto en normas con rango jerárquico superior en el ordenamiento jurídico. En cambio, la posibilidad de prescripción puede ser vista como una cualidad de los delitos que afecta a su pu...

Merleau-Ponty y la ontología de la naturaleza: intercorporalidad, negatividad y dialéctica

OpenAIRE

Basilio Cladakis, Maximiliano

2016-01-01

Resumen: El objetivo del siguiente trabajo es explicar cómo la elaboración de una ontología de la naturaleza por parte de Merleau-Ponty se circunscribe en el intento de elaborar un pensamiento del ser que no haga de éste un elemento del pensamiento abstracto, sino que dé cuenta de la experiencia concreta del ser en el mundo. En este sentido, son fundamentales los conceptos de intercorporalidad, negatividad y dialéctica. Abstract: The aim of the following work is to explain how the elaborat...

Concept of the LORELEI Test Device for LOCA Experiment in the JHR Reactor

International Nuclear Information System (INIS)

Moran, N.; Ferry, L.; Azulay, A.; Mileguir, O.; Weiss, Y.; Szanto, M.

2014-01-01

Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis. Former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. In JHR material testing reactor, which is currently under construction, one significant experimental device is the LORELEI testing device. The objective is to examine the LOCA sequence influence on: thermo-mechanical behavior of the fuel clad, possible fuel relocation, corrosion at high temperature, oxidation, hydriding and resulted clad embrittment. The device is a single rod closed loop system placed on a displacement device inside a defined channel in the reflector. Several operational constrains on the device, as required by the reactor operational philosophy resulted quite a few challenges in the design. Constrains as: pre experimental re-irradiation phase under thermo-syphonic flow, application of active insulation to simulate the surrounding fuel, application of tensile force during refolding simulation, controlling the experiment with non-direct temperature measurement, etc. requires sophisticated solutions. The main objective of the conceptual design was to remove the uncertainties of those challenging requirements. The current presentation describes the approach applied defining the concept of the device, using sophisticated design combined with computational and experimental tools

Hydrogen production in a PWR during LOCA

International Nuclear Information System (INIS)

Cassette, P.

1984-01-01

Hydrogen generation during a PWR LOCA has been estimated for design basis accident and for two more severe hypothetical accidents. Hydrogen production during design basis accident is a rather slow mechanism, allowing in the worst case, 15 days to connect a hydrogen recombining unit to the containment atmosphere monitoring system. Hydrogen generated by steam oxidation during more severe hypothetical accidents was found limited by steam availability and fuel melting phenomena. Uncertainty is, however, still remaining on corium-zirconium-steam interaction. In the worst case, calculations lead to the production of 500 kg of hydrogen, thus leading to a volume concentration of 15% in containment atmosphere, assuming homogeneous hydrogen distribution within the reactor building. This concentration is within flammability limits but not within detonation limits. However, hydrogen detonation due to local hydrogen accumulation cannot be discarded. A major uncertainty subsisting on hydrogen hazard is hydrogen distribution during the first hours of the accident. This point determines the effects and consequences of local detonation or deflagration which could possibly be harmful to safeguard systems, or induce missile generation in the reactor building. As electrical supply failures are identified as an important contributor to severe accident risk, corrective actions have been taken in France to improve their reliability, including the installation of a gas turbine on each site to supplement the existing sources. These actions are thus contributing to hydrogen hazard reduction

La educación ambiental y su importancia en la relación sustentable: Hombre-Naturaleza- Territorio

OpenAIRE

Ronald Fernando Quintana Arias

2017-01-01

Objetivo. Promover la importancia de la educación ambiental en el fortalecimiento de la relación sustentable Hombre-Naturaleza-Territorio en 8 niños y niñas observadores de aves de la primaria María Berchmans. Metodología. Entre agosto y noviembre del 2015 desarrollé un estudio de caso basado en el aprendizaje vivencial, social y experiencial, que se dividió en dos fases en las que se encontraban 19 actividades clasificadas dentro de siete...

BWR 200 % recirculation pump suction line break LOCA tests, RUNs 942 and 943 at ROSA-III without HPCS

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Koizumi, Yasuo

1986-03-01

This report presents the experimental results of RUNs 942 and 943 in ROSA-III program, which are 200 % recirculation pump suction line break LOCA tests with assumption of HPCS failure. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, ECCS's, and steam and feedwater systems. Effects of initial core void distribution and other fluid conditions on overall LOCA phenomena with special interest on transient core cooling phenomena were investigated by comparing the present test results with those of RUN 926, a 200 % suction line break test with standard initial fluid conditions. The initial core outlet quality was changed between 5 % and 43 %. As conclusions, (1) the initial lower core flow and higher void fraction affected significantly the core cooling conditions and resulted in earlier and higher PCT. (2) The lower plenum flashing temporarily contributed to cool down the core. (3) Flashing of remained hot water in the feedwater line affected slightly the pressure response and delayed the actuation of LPCI by 11 seconds. (4) The whole core was completely cooled down within 104 seconds after the LPCI actuation in these large break tests. (author)

Evaluation of PWR response to main-steamline break with concurrent steam-generator tube rupture and small-break LOCA

International Nuclear Information System (INIS)

Laaksonen, J.T.; Sheron, B.W.

1982-12-01

In 1980, the NRC staff raised a potential safety issue involving a coincident steamline break, steam generator tube rupture, and small-break loss-of-coolant accident (LOCA). The bases for this concern were that the system response, primarily the maintenance of core cooling, was unanalyzed and the adequacy of the present guidance to operators to respond to combination LOCAs was unknown. This report discusses the staff evaluations performed to assess the system response and the adequacy of the present emergency operator guidelines. In all of the analyzed cases the primary coolant shrinkage, caused by overcooling, and the simultaneous loss of coolant can be compensated by the high pressure emergency core cooling system. The core remains covered with liquid, and the primary coolant remains subcooled, except in the vessel upper head. If the steamline break is outside the containment and cannot be isolated, the radiological consequences could be more severe than in any accident currently analyzed in a typical plant Final Safety Evaluation Report (FSAR). To decrease the risk of elevated offsite releases, an early diagnosis of the tube rupture has to be ensured. This can be done by upgrading operator instructions. The appropriate mitigating actions are in the existing instructions

Mitigating Capability Analysis during LOCA for Korean Standard Nuclear Power Plants in Containment Integrity

Energy Technology Data Exchange (ETDEWEB)

Choi, Young; Park, Soo Yong; Kim, D. H.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2007-07-01

The objective of this paper is to establish Containment spray operational technical bases for the typical Korean Standard Nuclear Power plants (Ulchin units 3 and 4) by modeling the plant, and analyzing a loss of coolant accident (LOCA) using the MAAP code. The severe accident phenomena at nuclear power plants have large uncertainties. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand severe accident sequences and to assess the accident progression accurately by computer codes. Furthermore, it is important to attain the capability to analyze a advanced nuclear reactor design for a severe accident prevention and mitigation.

Reactor elements properties response during a postulated loss-of-coolant accident (LOCA)

International Nuclear Information System (INIS)

Ahmed, E.E.; Rahman, F.A.

1985-01-01

Four computer algorithms have been introduced to solve for the reactor different materials response subjected to LOCA conditions, they were developed with the intent of producing a simple, accurate and efficient prediction schemes. A general overview of the solution procedures design and working of each of four algorithms are presented, followed by short description of the nature of solution and calculated results. These algorithms are: 1. ZIRCP to give the cladding material properties response under normal and transient conditions. 2. FCGAPP to give the fuel- cladding gas-gap conductivity. 3. NFUEIP to solve the temperature dependent of nuclear fuel properties during normal and transient conditions. 4. TSDATP has been developed to solve for the thermodynamic and transport properties of water and steam over a large range of temperature and pressure. 14 fig

Estimation of the core-wide fuel rod damage during a LWR LOCA

International Nuclear Information System (INIS)

Mattila, L.; Sairanen, R.; Stengaard, J.-O.

1975-01-01

The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

Variaciones biopolíticas sobre naturaleza y vida

Directory of Open Access Journals (Sweden)

Espinosa Rubio, Luciano

2013-08-01

Full Text Available We need new narratives to understand our chaotic and highly complex world a little better. In this case it is useful to remember some historical models of the general relations between nature and life, that underlie biopolitics: a the traditional one, based on the natural law and life values; b the technical one, based on the will of power and the values of liberty; c the alternative one, based on the ecological cooperation between all living beings and both kinds of values. To survive in this crisis of civilisation, a lot of courage and solidarity is needed to confront the indifference that kills.Necesitamos nuevas narraciones para entender un poco más nuestro caótico y muy complejo mundo. En este caso es útil recordar algunos modelos históricos de las relaciones generales entre naturaleza y vida, que están en la base de la biopolítica: a el tradicional, basado en la ley natural y en los valores de la vida; b el técnico, basado en la voluntad de poder y en los valores de la libertad; c el alternativo, basado en la cooperación ecológica entre todos los seres vivos y en ambas clases de valores. Para sobrevivir en esta crisis de civilización hace falta mucho coraje y solidaridad frente a la indiferencia que mata.

Generating debate on issues surrounding the venting of containment in the event of LOCA to ensure an optimised safe outcome

International Nuclear Information System (INIS)

Chadwick, Chris; Jahouel, Xavier; Swain, Adam

2014-01-01

Following the incidents at the Japanese Fukushima Nuclear facility, when three units experienced LOCA, the consequences of those events have caused ripples across the world. Regulators around the world are examining the need to prevent excessive pressurisation of NPP Containment, and safely evacuate the gaseous consequences of LOCA, there is a need to return to fundamental principles and examine each putative set of data derived from the various models describing the consequence of LOCA and other events, a LOCA being a 'Loss Of Coolant Accident', and represents the outcome from a range of incidents ranging from fuel pellets being exposed to cooling (heat exchange) water, to a complete melt down of the fuel load with consequence evaporation of the concrete structures of the reactor containment buildings. Clearly it would be highly desirable in both economic and safety terms to minimise the external impact of these events inside the containment. Since one of the results of a LOCA is the pressurisation of the internal space of the containment building, regulators and the industry are looking at the mandatory installation of suitable vent systems to vent the pressure building up inside the containment, whilst ensuring the minimum impact on the surrounding environment. This is clearly a filtration/separation/recombination issue, and as an expert engineering company in the nuclear industry, Porvair has concerns that the issues of Containment Venting are not being addressed by expert filter companies, but by expert nuclear engineering companies with only a superficial knowledge of the complexities and nuances on filtration processes. The paper will describe in depth and detail the individual consequences of each particular aspect of the modelled data. Looking at flowing conditions (pressure, temperature, gas constituents including water vapour and Caesium and Iodine compounds), vent pressure philosophy, deposition of solids (size, type and quantity), decay heat

La apropiación de la naturaleza como recurso. Una mirada reflexiva

Directory of Open Access Journals (Sweden)

Gerardo Morales Jasso

2016-01-01

Full Text Available En el marco de la primera tarea de la historia ambiental, que es cuestionar la naturalidad aparente de la relación entre la sociedad y la naturaleza con el fin de repensar nuestro futuro, en este artículo se reflexiona sobre el concepto de recursos a partir de una epistemología sistémica, es decir, una epistemología no dualista. Por lo tanto, aunque abreva de las ciencias sociales, postula una crítica al concepto de recursos que en general se usa en éstas. Al cuestionar la aparente naturalidad del recurso, se pretende ampliar la categoría de “recursos” con el fin de que se generen investigaciones que permitan complementar y sustituir las perspectivas reduccionistas sobre uno de los problemas ambientales más acuciantes, el de los mal llamados “recursos naturales”.

OECD-LOFT large break LOCA experiments: phenomenology and computer code analyses

International Nuclear Information System (INIS)

Brittain, I.; Aksan, S.N.

1990-08-01

Large break LOCA data from LOFT are a very important part of the world database. This paper describes the two double-ended cold leg break tests LP-02-6 and LP-LB-1 carried out within the OECD-LOFT Programme. Tests in LOFT were the first to show the importance of both bottom-up and top-down quenching during blowdown in removing stored energy from the fuel. These phenomena are discussed in detail, together with the related topics of the thermal performance of nuclear fuel and its simulation by electric fuel rod simulators, and the accuracy of cladding external thermocouples. The LOFT data are particularly important in the validation of integral thermal-hydraulics codes such as TRAC and RELAP5. Several OECD partner countries contributed analyses of the large break tests. Results of these analyses are summarised and some conclusions drawn. 32 figs., 3 tabs., 45 refs

Hydrogen and steam distribution following a small-break LOCA in large dry containment

Institute of Scientific and Technical Information of China (English)

DENG Jian; CAO Xuewu

2007-01-01

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations.Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident (LOCA) without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump (RCP)compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.

¿Qué es en verdad el dinero? Una teoría sobre la naturaleza del dinero

OpenAIRE

Ramos-Arévalo, J.M. (Javier María)

2008-01-01

Sorprende descubrir que no hay una definición de dinero que sea admitida deforma universal. No hay una explicación de la naturaleza del dinero que justifique todas sus funciones y características. Este trabajo pretende demostrar que el dinero es ante todo un derecho, y que considerarlo un derecho nos puede permitir aclarar los enigmas del dinero. Existe bastante acuerdo sobre para qué surge el dinero, pero debemos preguntarnos el cómo, y a partir del ahí explicar de dónde le vi...

Calculation of BETHSY 0.5% small break LOCA with RELAP5-ISP 27 international activity of code assessment

International Nuclear Information System (INIS)

Chen Yuzhen

1992-01-01

BETHSY facility constructed in France is a 1/100 volumetrically-scaled full-pressure model of a PWR with 3 loops. ISP-27 is an international activity sponsored by OECD Nuclear Energy Agency. The experiment is a transient of 0.5% coldleg break LOCA with failure of HPIS. The calculations were performed with RELAP5/MOD2/36.05 at CYBER-170/825, which can present a good calculation, provided that the break flow is well modelled

La edad de la empatía. Lecciones de la naturaleza para una sociedad más justa y solidaria, por Fran de Waal

Directory of Open Access Journals (Sweden)

Cecilia Milito Barone

2015-07-01

Full Text Available Rsseña del libro Fran de Waal, La edad de la empatía. Lecciones de la naturaleza para una sociedad más justa y solidaria (Barcelona: Tusquets Editores, 2011, 258 pp. ISBN: 978-84-8383-350-6

Perfil del visitante de naturaleza en Latinoamérica: prácticas, motivaciones e imaginarios. Estudio comparativo entre México y Ecuador.

Directory of Open Access Journals (Sweden)

Maribel Osorio García

2017-01-01

Full Text Available El artículo expone los resultados de una investigación interinstitucional realizada sobre el com‑ parativo entre el perfil de los visitantes que llegan al Área de Protección de Flora y Fauna Nevado de Toluca, México y a la Reserva Ecológica Cotacachi Cayapas, Ecuador, ambas áreas naturales protegidas. Se toma como postura teórica que el comportamiento del visitante de naturaleza tiene una explicación no sólo a partir de sus motivaciones e interés ambiental, sino desde sus imaginarios. La metodología aplicada fue de carácter cuan‑ titativo a través de una encuesta realizada en cada sitio, considerando las motivaciones, el comportamiento de viaje, las prácticas ambientales y los imaginarios. Mediante un cruce de variables se caracterizaron cuatro perfiles: el recreativo, el deportivo, el de convivencia y el de contacto, siendo el primero el dominante. En cada caso se identifica el modo de interrelación que se establece con la naturaleza.

Lessons learned from OECD/CSNI ISP on small break LOCA: final report

International Nuclear Information System (INIS)

1996-07-01

This document presents an overview of the results obtained from five recent OECD/CSNI International Standard Problems (ISPs) dealing with phenomenologies typical of Small Break LOCA in PWR nuclear power plants of western design. The experiment in four Integral test Facilities, Lobi, Spes, Bethsy and Lstf and the recorded data from a steam generator tube rupture transient in the Belgian PWR of Doel, were taken as reference for the calculations. Relevant hardware characteristics of the facilities and of the plant are firstly given, including the correlation between key thermalhydraulic phenomena and the reference experimental scenarios. A statistical evaluation of the general data connected with each ISP is then presented. The lessons learned from the ISPs are then considered. Four areas have been identified: code deficiencies and capabilities, scaling of the data, progress in code capabilities and various additional aspects

Turismo de segundas residencias y turismo de naturaleza en el espacio rural mexicano

Directory of Open Access Journals (Sweden)

David Vargas del Río

2015-01-01

Full Text Available La posmodernidad y los imaginarios urbanos son aspectos que contribuyen a la hegemonía de la ciudad sobre el campo. El turismo es un agente fundamental en este proceso y dos aspectos importantes son el turismo de segundas residencias y el turismo basado en la naturaleza; sin embargo, la relación entre ambos no es clara. El concepto de la urbanización del campo se utiliza como fundamento para describir las condiciones materiales y culturales que facilitaron el avance de estas dos formas de turismo en el espacio rural mexicano. Después, mediante un estudio transversal referido a subzonas de la geografía local de Mazunte, Oaxaca, se ilustra la forma como ambos fenómenos se articularon geográficamente mientras se demuestra su estrecha relación.

Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

International Nuclear Information System (INIS)

Gonzalez, Jose; Gimenez, Marcelo

2000-01-01

We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

A 25% double-ended LOCA in the PSB-VVER facility

International Nuclear Information System (INIS)

Blinkov, V.N.; Lipatov, I.A.; Dremin, G.I.

2003-01-01

The paper presents the results of the experimental investigation in the PSB-VVER facility and post-test analysis with RELAP5/MOD3.3 of the thermal-hydraulic response of the PSB-VVER to a 25% double-ended Hot Leg Break (HLB). The test scenario included loss of off-site power concurrently with the scram-signal and the safety system operation as described in the reference VVER-1000 operational manual in the case of this type of accident assuming one diesel-generator failure. The key transient parameter trends as well as sequence of events and phenomena are given in the paper. RELAP5/MOD3.3 a post-test analysis has been performed using the experimental data gained as a base. The reasonable qualitative agreement between the key calculated and measured variables has been shown. The quantitative code accuracy evaluation has shown that the total average amplitude of the main parameters' deviations AA tot tot < 0.28 that corresponds to satisfactory quality of the VVER-1000 hot leg guillotine break LOCA modeling in the PSB-VVER. (author)

MAAP-CANDU simulations of LOCA/LOECI accidents at Darlington NGS

International Nuclear Information System (INIS)

Kwee, M.T.; Choi, M.H.; Leung, R.K.

1996-01-01

Severe accidents have been the subject of a great deal of analysis and research, particularly in the light water reactor community. Although severe accident analysis in Canada deuterium-uranium (CANDU) reactors has not been published abundantly, a significant body of research and analysis has been accumulated. This has occurred because CANDU has directly taken into consideration a set of severe accidents [e.g loss-of-coolant accidents (LOCAs) coincident with a loss-of-emergency-coolant injection (LOECI)] in the design basis. These accidents have served to define the design requirements that ensure the integrity of the heat transport system. The CANDU reactor design has inherent heat sinks such as the primary heat transport system, the secondary side, moderator system, and shielding system (shield tank and end shields). These heat sinks are significant and are able to moderate or terminate the progression of severe accidents that go beyond the design base cases. These types of accidents are typically analyzed at Ontario Hydro in conjunction with probabilistic safety analysis (PSA), where the severe accident consequences are analyzed by a series of conservative hand-calculation methods

Strain measurements at the HDR-pipe-system under LOCA-load: Effects on elbows and displaced weldings

International Nuclear Information System (INIS)

Hunger, H.

1985-01-01

This paper characterizes some effects which have been detected during strain gauge measurements on a test piping with feed water check valve oscillating under blowdown-load. The ovalization of a pipe elbow subjected to in-plane-bending affects the connected straight pipe; this is shown by means of circumferential stresses. Very high LOCA-load produces plastic strain and changes the pipe dynamics. Artificial displaced welds increase the local strain but no defects have occurred. One example compares stresses from measurement and post-calculation. Moreover there are given some remarks on the optimization of the comparison of measurement and calculation. (orig.)

Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

International Nuclear Information System (INIS)

Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

1988-12-01

A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

Género, naturaleza y política: Los estudios sobre género y medio ambiente

OpenAIRE

Diana Ojeda Ojeda

2011-01-01

El campo intelectual conocido como género y medio ambiente estudia las complejas dinámicas de poder que entretejen los procesos simultáneos de producción de la naturaleza y producción de sujetos, prestando especial atención a cómo dichos sujetos y su relación con el entorno están mediados por procesos de diferenciación y dominación basados en género. Este artículo hace un balance de esta literatura a partir de tres ejes y aboga por una historia ambiental que logre comprender mejor el papel qu...

CONTEMPT-4MOD3, LWR Containment Long-Term Pressure Distribution and Temperature Distribution in LOCA

International Nuclear Information System (INIS)

Lin, C.C.; Economos, C.; Lehner, J.R.; Maise, G.; Ng, K.K.; Mirsky, S.M.

2002-01-01

1 - Description of problem or function: CONTEMPT-4/MOD5 describes the response of multi-compartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user- supplied descriptions of compartments, inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. To accommodate degraded core type accidents, analytical models for hydrogen combustion within compartments and energy transfer due to gas radiation are also provided. CONTEMPT4/MOD6 is an update of previous CONTEMPT4 versions. Improvements in CONTEMPT4/MOD6 over CONTEMPT4/MOD3 include coding of a BWR pressure suppression system model, a hydrogen/carbon monoxide burn model, a gas radiation heat transfer model, a user specified variable junction (leakage) area as a function of pressure or time, additional heat transfer coefficient options for heat structures, generalized initial compartment conditions for inerted containment, an alternative containment spray model and spray carry-over capability. Also, the thermodynamic properties routines have been extended to accommodate the higher temperature and multicomponent gas mixtures associated with combustion. In addition, reduced running time is achieved by incorporation of an optional implicit numerical algorithm for junction flow. This makes economically feasible the analysis of very long

Deshidratación de etanol empleando líquidos iónicos de naturaleza prótica

OpenAIRE

Acosta Cordero, L.; Pérez Ones, O.; Zumalacárregui de Cárdenas, L.

2017-01-01

En este trabajo se presenta el estudio experimental de la deshidratación de etanol empleando líquidos iónicos de naturaleza prótica, con el objetivo de determinar un líquido iónico prótico efectivo en este proceso de separación. Para ello se sintetizaron tres líquidos iónicos (formiato de 2-hidroxietilamonio, lactato de2-hidroxietilamonio y propionato de 2-hidroxietilamonio) y se determinaron sus propiedades críticas y densidad apartir de métodos de contribución de grupos. Además se confirmó ...

Biodiversidad exportada y regiones transformadas: naturaleza, comercio y dinámica regional en Costa Rica (1884-1948)

OpenAIRE

Anthony Goebel Mc Dermott

2014-01-01

El presente artículo tiene como objetivos centrales, por un lado, visibilizar, desde una perspectiva socioambiental, y valiéndose de referentes teóricos como la Economía Ecológica, el papel de los “bienes” de explotación extractiva –con especial atención a los recursos naturales bióticos- en la inserción directa de la naturaleza “costarricense” en el mercado mundial, y por otro, dar cuenta del carácter regionalmente diferenciado de la presión económica de este último sobre los diversos ecosis...

La naturaleza interior. El árbol como referente simbólico en la arquitectura contemporánea japonesa

OpenAIRE

López del Río, Alberto

2013-01-01

El análisis planteado en el presente trabajo pretende dar, en cierta medida, una explicación a algunas de las soluciones más sorprendentes de la arquitectura japonesa de finales del siglo XX y principios del XXI, que tienen en el uso del árbol como referente simbólico un denominador común, a la vez que se busca esclarecer las motivaciones y referentes del empleo del mismo en estos proyectos. Como queda patente, la relación con la naturaleza es uno de los conceptos claves de la cultura japo...

Selección de las formas matemáticas en la naturaleza y su emergencia en la arquitectura

OpenAIRE

Torrens Bermejo, Mercedes

2013-01-01

[ES]El trabajo analiza el empleo en el diseño arquitectónico de las formas matemáticas presentes en la naturaleza . Examina las formas de los objetos resultantes la selección fundamental, de la natural y de la cultural. Conjetura sobre el arte como una consecuencia de la evolución del ser humano al transformar la realidad partiendo de las formas inteligibles y/o bellas que percibe en su entorno. [EN] The paper analyzes the use in the architectural design of the mathematical forms fou...

Educación ambiental: itinerario en la naturaleza y su relación con conectividad, preocupaciones ambientales y conducta

OpenAIRE

Olivos-Jara, Pablo; Aragonés, Juan-Ignacio; Navarro-Carrascal, Oscar

2013-01-01

Resumen Las excursiones e itinerarios a través de entornos naturales son frecuentes en la educación ambiental, pero existe poca evidencia empírica de investigaciones que aporten resultados que permitan comprender mejor los efectos de las experiencias de contacto directo con ambientes naturales, y menos aún acerca de su relación con la conectividad con la naturaleza, las preocupaciones ambientales y el comportamiento proambiental. Se llevan a cabo dos estudios, en el primero participan 286 est...

Educación ambiental: itinerario en la naturaleza y su relación con conectividad, preocupaciones ambientales y conducta

OpenAIRE

Pablo Olivos-Jara; Juan-Ignacio Aragonés; Oscar Navarro-Carrascal

2013-01-01

Las excursiones e itinerarios a través de entornos naturales son frecuentes en la educación ambiental, pero existe poca evidencia empírica de investigaciones que aporten resultados que permitan comprender mejor los efectos de las experiencias de contacto directo con ambientes naturales, y menos aún acerca de su relación con la conectividad con la naturaleza, las preocupaciones ambientales y el comportamiento proambiental. Se llevan a cabo dos estudios, en el primero participan 286 estudiantes...

Verification of human actions in SBO sequences with LOCA stamps in Westinghouse PWRs; Verificacion de las actuaciones humanas en secuencias de SBO con LOCA de sellos en reactores PWR Westinghouse

Energy Technology Data Exchange (ETDEWEB)

Queral, C.; Mena Rosell, L.; Jimenez Varas, G.

2013-07-01

The Fukushima accident has shown the need for tools and methodologies able to analyze human activities and / or capabilities of portable systems that has given the Spanish plants as a result of the stress tests . In this work we have applied the methodology of integrated safety analysis developed by the CSN , to SBO sequences with LOCA stamp. The aim is to show a methodology for testing the performances of the Emergency Operating Procedures and Guides Severe Accident Management. The simulations were performed with the tool SCAIS coupled to MAAP . The results show that there are human activities that may be beneficial in certain sequences but harmful in others. This type of problem is already known and referred to in the GGAS . However, FSR shows a practical way to check human actions cannot be obtained with other methods.

KUPOL-M code for simulation of the VVER's accident localization system under LOCA conditions

International Nuclear Information System (INIS)

Efanov, A.D.; Lukyanov, A.A.; Shangin, N.N.; Zajtsev, A.A.; Solov'ev, S.L.

2004-01-01

Computer code KUPOL-M is developed for analysis of thermodynamic parameters of medium within full pressure containment for NPPs with VVER under LOCA conditions. The analysis takes into account the effects of non-stationary heat-mass transfer of gas-drop mixture in the containment compartments with natural convection, volume and surface steam condensation in the presence of noncondensables, heat-mass exchange of the compartment atmosphere with water in the sumps. The operation of the main safety systems like a spray system, hydrogen catalytic recombiners, emergency core cooling pumps, valves and a fan system is simulated in KUPOL-M code. The main results of the code verification including the ones of the participation in ISP-47 International Standard Problem on containment thermal-hydraulics are presented. (author)

Naturaleza y origen de los demonios en la épica tibetana de Gesar de Ling

Directory of Open Access Journals (Sweden)

Bertha Aceves Torres

2013-05-01

Full Text Available El poema épico del héroe tibetano Gesar de Ling nace de tradición oral, posteriormente es escrito en diversas lenguas, y sus interpretaciones orales, en el Tíbet, perduran hasta nuestros días. El conflicto que desata las acciones en el relato es el enfrentamiento del héroe contra los seres demoníacos, que asolan y subyugan al pueblo tibetano. La categoría de demonios en el occidente difiere de la naturaleza de los demonios en el budismo tibetano. La subyugación de los demonios por Gesar de Ling, que se efectúa en el poema, corresponde a uno de los objetivos del budismo a su llegada al Tíbet.

VVER-1000 small-medium break LOCAs predictions by ASTEC

International Nuclear Information System (INIS)

Georgieva, J.; Stefanova, A.; Atanasova, B.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

2005-01-01

This paper deals with an assessment of ASTEC1.1v0 code in the simulation of small and medium break LOCAs (ranging from 30mm up to 70mm equivalent diameter). The reference power plant for this analysis is a VVER-1000/V320 (e.g. Units 5 and 6 at Kozloduy NPP). A preliminary comparison with MELCOR and RELAP-SCDAP severe accident codes will be discussed. This investigation has been performed in the framework of the SARNET project (under the Euratom 6 th framework program) by the FoBAUs group (Forum of Bulgarian ASTEC users). The FoBAUs group aims at the validation of the ASTEC code in the field of severe accidents. Future activities will target the ASTEC capability (as a PSA-level 2 tool) to simulate a large range of reactor accident scenarios with intervention of safety systems (either passive systems or operated by operators). The final target is to assess Severe Accident Management (SAM) procedures for VVER-1000 reactors. The ASTEC1.1v0 code version here used is the one released in June 2004 by the French IRSN (Institut de Radioprotection et de Surete Nucleaire) and the German GRS (Gesellschaft ReactorSicherheit mbH). (author)

HCCR TBS LOCA and ICE into small confined volume

International Nuclear Information System (INIS)

Jin, Hyung Gon; Ahn, Mu-Young

2016-01-01

KAERI has participated in the development of HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) as a member of the KO TBM Team. Conceptual design review of this system had been performed in 2015 and after resolving the chits, the final approval was achieved in March 2016. This safety issue is one of the category II chits in the CDR and resolution strategy was already approved, however, safety analysis should be done until PDR (Preliminary Design Review). In this paper, model and nodalization for the accident are given and preliminary result is included. Nominal design pressure of HCS loop is 8 MPa, therefore, as indicated in the figure below. During the break of cooling pipe between TBM and Shield, the high pressure coolant will ingress to the 'interspace' between TBM, Shield and Frame. The coolant will be released through the front gaps between TBM and Frame towards VV primary vacuum. Accident analysis about HCCR TBS LOCA and ICE into small confined volume has been done successfully. Inverspace volume is compatibly small volume for 8MPa helium loop rupture, which causes fast pressure build-up the space but it decrease within 10 seconds. It is expected that other type of TBM has almost the same behavior

THE PREDICTION OF pH BY GIBBS FREE ENERGY MINIMIZATION IN THE SUMP SOLUTION UNDER LOCA CONDITION OF PWR

Directory of Open Access Journals (Sweden)

HYOUNGJU YOON

2013-02-01

Full Text Available It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3&4 and UCN 1&2. As results, pH of the sump solution for the SKN 3&4 was between 7.02 and 7.45, and for the UCN 1&2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl, HNO3, and Cs are very low.

Verification of human actions in SBO sequences with LOCA stamps in Westinghouse PWRs

International Nuclear Information System (INIS)

Queral, C.; Mena Rosell, L.; Jimenez Varas, G.

2013-01-01

The Fukushima accident has shown the need for tools and methodologies able to analyze human activities and / or capabilities of portable systems that has given the Spanish plants as a result of the stress tests . In this work we have applied the methodology of integrated safety analysis developed by the CSN , to SBO sequences with LOCA stamp. The aim is to show a methodology for testing the performances of the Emergency Operating Procedures and Guides Severe Accident Management. The simulations were performed with the tool SCAIS coupled to MAAP . The results show that there are human activities that may be beneficial in certain sequences but harmful in others. This type of problem is already known and referred to in the GGAS . However, FSR shows a practical way to check human actions cannot be obtained with other methods.

Presentación. Género, cuerpo y sexualidad. Cultura y ¿Naturaleza?

Directory of Open Access Journals (Sweden)

AIBR. Consejo de Redacción

2006-01-01

Full Text Available En 1974, se recogió en un libro editado por Michelle Zimbalist Rosaldo y Louise Lamphere, las ponencias sobre género que habían sido presentadas en una mesa sobre “mujeres, hombres, cultura y sociedad” en el Congreso de la American Anthropological Association de 1972, celebrado en Toronto. De ellas, las tres primeras tuvieron una gran resonancia por su intento de explicar el universal de la dominación masculina. Michelle Rosaldo (1974 planteaba un modelo estructural que aunaba aspectos de la psicología, la organización social y cultural relacionándolos con la oposición entre lo público y lo privado. Sherry Ortner (1974, con una orientación influenciada por el estructuralismo de Lévi-Strauss, estudiaba los paralelismos entre las dicotomías hombre / mujer y naturaleza / cultura; mientras Nancy Chodorow (1974 se centraba en las relaciones y estructura familiar desde una perspectiva más psicológica con influencia de los planteamientos freudianos.

TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test

International Nuclear Information System (INIS)

Richner, M.; Analytis, G.Th.; Aksan, S.N.

1993-01-01

1 - Description of test facility: NEPTUN is designed to perform PWR LOCA simulation experiments, which provide the full length emergency cooling heat transfer tests for LOFT. Therefore the NEPTUN heater bundle with 33 electrical heater elements and 4 guide tubes simulates a section of the LOFT nuclear core. The main test loop also contains measuring systems for the carry-over rate and for the steam expelled, and a back-pressure control system. A water loop brings the water to the initial reflooding conditions. In addition, auxiliary systems maintain normal operating conditions. 2 - Description of test: Test 5052 is one of a series of 40 reflood tests performed in NEPTUN. Before the start of the test, the flooding water in its circuit is brought to the following conditions: pressure = 4.1 bar; velocity = 2.5 cm/sec; subcooling temperature = 78 C; single rod power = 2.45 kW; maximal initial cladding temperature = 867 C. 3 - Status: CSNI1013/01, 21-Jul-1993 Arrived at NEADB

A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

Directory of Open Access Journals (Sweden)

Jae-Yong Kim

2014-01-01

Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

International Nuclear Information System (INIS)

Bujan, A.; Adamik, V.; Misak, J.

1986-01-01

A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

Propuesta para la enseñanza de Ley de Coulomb contemplando aspectos de la naturaleza de las ciencias

OpenAIRE

Castellanos Clavijo, Darío Eusebio

2014-01-01

La propuesta para la Enseñanza De Ley De Coulomb se materializa en una unidad didáctica que contempla un discurso pedagógico que se ha interesado por la enseñanza de las ciencias naturales en este caso la física, denominado La naturaleza de las ciencias, discurso que se ha preocupado en las últimas décadas en posibilitar al estudiante el conocimiento de la ciencia desde lo que la caracteriza, como se produce y para que se produce. A su vez la unidad integra estándares de competencia par...

A novel procedure to measure the antioxidant capacity of Yerba maté extracts Procedimento padronizado para avaliar a capacidade antioxidante dos extratos de erva-mate

Directory of Open Access Journals (Sweden)

Vanessa Graciela Hartwig

2012-03-01

Full Text Available Yerba maté extracts have in vitro antioxidant capacity attributed to the presence of polyphenolic compounds, mainly chlorogenic acids and dicaffeoylquinic acid derivatives. DPPH is one of the most used assays to measure the antioxidant capacity of pure compounds and plant extracts. It is difficult to compare the results between studies because this assay is applied in too many different conditions by the different research groups. Thus, in order to assess the antioxidant capacity of yerba maté extracts, the following procedure is proposed: 100 µL of an aqueous dilution of the extracts is mixed in duplicate with 3.0 mL of a DPPH 'work solution in absolute methanol (100 µM.L-1, with an incubation time of 120 minutes in darkness at 37 ± 1 °C, and then absorbance is read at 517 nm against absolute methanol. The results should be expressed as ascorbic acid equivalents or Trolox equivalents in mass percentage (g% dm, dry matter in order to facilitate comparisons. The AOC of the ethanolic extracts ranged between 12.8 and 23.1 g TE % dm and from 9.1 to 16.4 g AAE % dm. The AOC determined by the DPPH assay proposed in the present study can be related to the total polyphenolic content determined by the Folin-Ciocalteu assay.Extratos de erva-mate têm a sua capacidade antioxidante in vitro atribuída à presença de compostos polifenólicos, principalmente ácidos clorogênicos e derivados do ácido dicafeoilquínico. Embora DPPH seja um dos ensaios mais utilizados para avaliar a capacidade antioxidante dos compostos puros e extratos de plantas, o fato de que há uma padronização pobre na sua aplicação torna as comparações entre os diferentes extratos muito difíceis. Visando conseguir uma técnica padronizada para medir a capacidade antioxidante de extratos de erva-mate, propomos o seguinte procedimento: 100 µL de uma diluição do extrato aquoso são misturados em duplicata, com 3,0 mL de uma solução de trabalho de DPPH em metanol absoluto

The Effect of External Vessel Cooling for a 2 inch LOCA Severe Accident Scenario at SMART with MIDAS/SMR

International Nuclear Information System (INIS)

Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

2010-01-01

KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. This feature may prevent the large size of LOCA. However it is necessary to estimate the hypothetical severe accidents progression for improving the degree of safety and identifying the unknown weakness of the system against an accident. To simulate a hypothetical severe accident for the SMART, we adopt the MIDAS/SMR code which was developed by KAERI

Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

Energy Technology Data Exchange (ETDEWEB)

Barrio del Juanes, M T; Garcia Cuesta, J C; Vallejo Diaz, I; Puebla, Herranz

2001-07-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

Energy Technology Data Exchange (ETDEWEB)

Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

2001-07-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

Primer registro para la flora argentina de Ilex affinis (Aquifoliaceae, sustituto de la "yerba mate" First report for Argentina of Ilex affinis (Aquifoliaceae, a "mate" substitute

Directory of Open Access Journals (Sweden)

Héctor A. Keller

2011-06-01

Full Text Available El hallazgo de Ilex affinis Gardner (Aquifoliaceae en los paredones rocosos y áreas pantanosas del paraje Teyú Cuaré, Misiones, Argentina, permite elevar a siete el número de especies de este género para nuestro país y desplaza al sur el límite austral de dispersión de esta especie. Se describe la especie sobre la base de los ejemplares hallados, y se la ilustra mediante fotografías. Además se presentan consideraciones sobre su distribución, observaciones ecológicas y su relación con la yerba mate (Ilex paraguariensis A. St. Hil..The finding of Ilex affinis Gardner (Aquifoliaceae in the rocky cliffs and swampy places of Teyú Cuaré, Misiones Province, Argentina, raises to seven the number of species of this genus for our country and moves southward the distribution limits of the species. The species is described on the basis of the specimens collected, and it is illustrated by photographies. Moreover, considerations about distribution, ecology and relationships with maté (Ilex paraguariensis A. St. Hil. are given.

Prediction of the fuel failure following a large LOCA using modified gap heat transfer model

International Nuclear Information System (INIS)

Lee, K.M.; Lee, N.H.; Huh, J.Y.; Seo, S.K.; Choi, J.H.

1995-01-01

The modified Ross and Stoute gap heat transfer model in the ELOCA.Mk5 code for CANDU safety analysis is based on a simplified thermal deformation model. A review on a series of recent experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this study, more realistic offset crap conductance model is implemented in the code to estimate the fuel failure thresholds usincr the transient conditions of a 100% Reactor Outlet Header (ROH) break LOCA. Based on the offset gap conductance model, the total release of I-131 from the failed fuel elements in the core is reduced from 3876 TBq to 3283 TBq to increase margin for dose limit. (author)

Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

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Lázaro Lima

2011-12-01

Full Text Available “Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

Naturaleza y ciudad en épocas prehispánicas. El caso de las culturas nucleares sudamericanas

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Leonel Ramos Santibáñez

2007-11-01

Este trabajo, busca dilucidar algunos de estos temas en el particular caso sudamericano prehispánico, donde la enorme cantidad de frágiles e interrelacionados ecosistemas que alberga la variedad geográfica, unida a su inestabilidad climática y geológica, hizo que el hombre desarrollara a lo largo de miles de años, un conocimiento y un pensamiento dirigido a encontrar los medios tecnológicos necesarios para buscar la forma de integrarse armónicamente a la Naturaleza. Para esto, a diferencia de otras culturas, desarrolló el particular modo de ver y entender el mundo como un todo vivo e interrelacionado del que el hombre es parte indesligable, que se muestra en este trabajo.

La relación entre las dicotomías cultura-naturaleza, hombre-mujer y humano-animal en el pensamiento feminista

OpenAIRE

Rodríguez Carreño, Jimena

2016-01-01

Gran parte del pensamiento feminista ha combatido el dualismo hombre/mujer al tiempo que ha tolerado el que enfrenta a la naturaleza con la cultura y a los humanos con los animales. Lo cierto es, sin embargo, que estos tres dualismos se encuentran tan vinculados que, para combatir uno de ellos, hay que combatirlos todos. Mostrar esta relación esencial entre dichas dicotomías es el objetivo fundamental de este trabajo. Al analizar tales dualismos, surgen otros también de gran importan...

Influence of partial blockage of a BWR bundle on heat transfer, cladding temperature, and quenching during bottom flooding or top spraying under simulated LOCA conditions

International Nuclear Information System (INIS)

Brand, B.; Gaul, H.P.; Sarkar, J.

1982-01-01

In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m 3 /h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800 0 C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis

Ecofeminismo. Una reivindicación de la mujer y la naturaleza

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María Tardón Vigil

2011-06-01

Full Text Available A lo largo de la historia, la mujer ha sido, entre otras muchas cosas, creadora de conocimientos, productora de materia y guardiana de la biodiversidad. Los rasgos androcéntricos han ido transformando la cultura hasta desembocar en una crisis ecológica, razón de más por la cual se aborda la cuestión medioambiental desde el lado femenino (resaltando los valores de la naturaleza, la mujer, el animal, el sentimiento, la materia y el cuerpo. Fue Françoise d’Eaubonne la primera que usó el término “ecofeminismo” reclamando el cuerpo femenino como propiedad de una misma. Y a partir de esta premisa, diversos grupos de mujeres comenzaron a tomar conciencia sobre los riesgos de su salud, derivados del uso de pesticidas, fertilizantes y excesiva medicalización que repercuten sobre el cuerpo femenino. Y ya Karen Warren hace ecofeminismo desde la filosofía, concibe la diferenciación como inferioridad. La dominación es la peor de las formas de maltrato, y el medioambiente sufre cada día las consecuencias, cierta reflexión habría de impulsar una compasión especial por ese otro que nunca protesta.

Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

International Nuclear Information System (INIS)

Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

2012-01-01

An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs

International Nuclear Information System (INIS)

Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

2013-01-01

SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

FORMACIÓN DEL PROFESORADO DE CIENCIAS Y ENSEÑANZA DE LA NATURALEZA DE LA CIENCIA

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José Antonio Acevedo Díaz

2010-09-01

Full Text Available La formación del profesorado para la implementación de una enseñanza de la naturaleza de la ciencia (NdC con calidad es un aspecto clave de la didáctica de las ciencias actual. En este artículo, se exponen los componentes principales del Conocimiento Didáctico del Contenido (CDC para la enseñanza de la NdC, como el conocimiento base que debe conseguir un profesor de ciencias en su desarrollo profesional. Se proponen un modelo integrador del CDC-NdC, varios contextos parapresentar la NdC en el aula y el uso de un enfoque explícito y reflexivo. La formación del profesorado de ciencias debe prestar atención a los aspectos indicados y la didáctica de la ciencia investigarlos.

FORMACIÓN DEL PROFESORADO DE CIENCIAS Y ENSEÑANZA DE LA NATURALEZA DE LA CIENCIA

Directory of Open Access Journals (Sweden)

José Antonio Acevedo Díaz

2010-01-01

Full Text Available La formación del profesorado para la implementación de una enseñanza de la naturaleza de la ciencia (NdC con calidad es un aspecto clave de la didáctica de las ciencias actual. En este artículo, se exponen los componentes principales del Conocimiento Didáctico del Contenido (CDC para la enseñanza de la NdC, como el conocimiento base que debe conseguir un profesor de ciencias en su desarrollo profesional. Se proponen un modelo integrador del CDC-NdC, varios contextos para presentar la NdC en el aula y el uso de un enfoque explícito y reflexivo. La formación del profesorado de ciencias debe prestar atención a los aspectos indicados y la didáctica de la ciencia investigarlos.

Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs; Comparacion de la fenomenologia de las secuencias de SBO con y sin LOCA de sellos en reactores PWR Westinghouse

Energy Technology Data Exchange (ETDEWEB)

Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

2013-07-01

SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

Construcción de demandas políticas de comunidades Mbyá guaraníes en contextos de conservación de la naturaleza

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Muriel Papalia

2012-12-01

Full Text Available A partir del análisis de un conflicto que enfrentó a comunidades Mbyá y empresas forestales en la Reserva de Biosfera Yabotí (Misiones, se evidenció cómo referencias a la conservación de la naturaleza son recuperadas por estas comunidades para energizar la construcción de demandas políticas orientadas a activar la representación indígena en el "comité de gestión" de la Reserva de Biosfera Yabotí. Las comunidades profundizan una "estructura de oportunidad política" entablando alianzas con sectores ambientalistas e indigenistas e interpelando a la red de vínculos organizada en torno a esta categoría de conservación promovida por la Unesco. El proceso analizado evidencia procesos políticos donde se tensionan prácticas y categorías sociales relativas al rol de los sujetos indígenas en iniciativas de gestión "participativas" de la naturaleza, señalando además cómo, en este proceso, las comunidades generan modos de acción política ambientalmente informada.

Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

International Nuclear Information System (INIS)

Kolev, N.I.; Sybotinov, L.S.

1984-01-01

The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

Analytical methods of leakage rate estimation from a containment under a LOCA

International Nuclear Information System (INIS)

Chun, M.H.

1981-01-01

Three most outstanding maximum flow rate formulas are identified from many existing models. Outlines of the three limiting mass flow rate models are given along with computational procedures to estimate approximate amount of fission products released from a containment to environment for a given characteristic hole size for containment-isolation failure and containment pressure and temperature under a loss of coolant accident. Sample calculations are performed using the critical ideal gas flow rate model and the Moody's graphs for the maximum two-phase flow rates, and the results are compared with the values obtained from then mass leakage rate formula of CONTEMPT-LT code for converging nozzle and sonic flow. It is shown that the critical ideal gas flow rate formula gives almost comparable results as one can obtain from the Moody's model. It is also found that a more conservative approach to estimate leakage rate from a containment under a LOCA is to use the maximum ideal gas flow rate equation rather than the mass leakage rate formula of CONTEMPT-LT. (author)

Habitar la naturaleza en armonía con el universo. Metafísica, geometría cósmica y orden social en la tradición arquitectónica china

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José Manuel Almodóvar Melendo

2016-01-01

Full Text Available En este artículo se establece un paralelismo entre la arquitectura tradicional china y concepciones cosmológicas y sociales, con objeto de analizar su proceso de formalización. En este sentido, se discute el modo en que los patrones arquitectónicos se han estandarizado a partir de módulos o elementos relacionados en diferentes niveles de composición y articulados en torno al vacío, que simboliza la naturaleza, para alcanzar el equilibrio con el Cosmos. Las conclusiones evidencian que se han establecido unas relaciones llenas de significado entre el hombre y la naturaleza, que responden simultáneamente a una síntesis de ideologías y filosofías jerarquizadas que provienen del confucianismo y su continuo desarrollo.

Ethylene propylene cable degradation during LOCA research tests: tensile properties at the completion of accelerated aging

International Nuclear Information System (INIS)

Bustard, L.D.

1982-05-01

Six ethylene-propylene rubber (EPR) insulation materials were aged at elevated temperature and radiation stress exposures common in cable LOCA qualification tests. Material samples were subjected to various simultaneous and sequential aging simulations in preparation for accident environmental exposures. Tensile properties subsequent to the aging exposure sequences are reported. The tensile properties of some, but not all, specimens were sensitive to the order of radiation and elevated temperature stress exposure. Other specimens showed more severe degradation when simultaneously exposed to radiation and elevated temperature as opposed to the sequential exposure to the same stresses. Results illustrate the difficulty in defining a single test procedure for nuclear safety-related qualification of EPR elastomers. A common worst-case sequential aging sequence could not be identified

Assessment on the Reactor Containment Cooling Capability of Kori Unit 1 Under LOCA Conditions with Loss of Offsite Power

International Nuclear Information System (INIS)

Lee, Jin Yong; Park, Jong Woon; Kim, Hyeong Taek

2006-01-01

The fan cooler system is designed to remove heat from containment under postulated accident conditions. During a postulated LOCA concurrent with a Loss of Offsite Power (LOOP), the Component Cooling Water (CCW) pumps that supply cooling water to the fan cooler and the fan that supplies containment air to the fan cooler will temporarily lose power. Then, the high temperature steam in the containment atmosphere will pass over the fan cooler tubing without forced cooling water flow. In that case, boiling may occur in the fan cooler tubes causing steam bubbles to form and pass into the attached CCW piping creating steam voids. Prior to the CCW pumps restart, the presence of steam and subcooled water can induce the potential for water hammer. As the CCW pumps restart, the accumulated steam condenses and the pumped water can produce a water hammer when the void closes. The hydrodynamic loads caused by such a water hammer event could challenge the integrity and the function of the fan cooler and associated CCW system. With respect to this phenomena, the United States Nuclear Regulatory Commission (USNRC) issued the Generic Letter (GL) 96-06, which requests an assessment of the possibility of boiling and water hammer in the cooling water system. The objectives of this study are to develop a analysis method for predicting the thermal hydraulic status of containment fan cooler and then to assess the containment fan cooler of Kori Unit 1 using the developed model under a LOCA with LOOP

Naturaleza fractal en redes de cristales de grasas

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Gómez Herrera, C.

2004-06-01

Full Text Available The determination of the mechanical and rheological characteris­tics of several plastic fats requires a detailed understanding of the microstructure of the fat crystal network aggregates. The (or A fractal approach is useful for the characterization of this micros­tructure. This review begins with information on fractality and statistical self-similar structure. Estimations for fractal dimension by means of equations relating the volume fraction of solid fat to shear elastic modulus G' in linear region are described. The influence of interesterification on fractal dimension decrease (from 2, 46 to 2 ,15 for butterfat-canola oil blends is notable . This influence is not significant for fat blends without butterfat. The need for an increase in research concerning the relationship between fractality and rheology in plastic fats is emphasized.La determinación de las características mecánicas y reológicas de ciertas grasas plásticas requiere conocimientos detallados sobre las microestructuras de los agregados que forman la red de cristales grasos. El estudio de la naturaleza fractal de estas microestructuras resulta útil para su carac­terización. Este artículo de información se inicia con descripciones de la dimensión fractal y de la "autosimilitud estadística". A continuación se describe el cálculo de la dimensión fractal mediante ecuaciones que relacionan la fracción en volumen de grasa sólida con el módulo de recuperación (G' dentro de un comportamiento viscoelástico lineal. Se destaca la influencia que la interesterificación ejerce sobre la dimensión fractal de una mezcla de grasa láctea y aceite de canola (que pasa de 2,64 a 2,15. Esta influencia no se presenta en mezclas sin grasa láctea. Se insiste sobre la necesidad de incrementar las investi­gaciones sobre la relación entre reología y estructura fractal en grasas plásticas.

La selección de contenidos para enseñar naturaleza de la ciencia y tecnología (parte 2: Una revisión desde los currículos de ciencias y la competencia PISA

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Ángel Vázquez-Alonso

2012-01-01

Full Text Available Este estudio revisa los currículos de ciencias (con especial énfasis en el español y la competencia científica de PISA como fuentes de contenidos de aprendizaje correspondientes a la "naturaleza de la ciencia" (y tecnología. Este es continuación de la primera parte donde se expusieron los consensos de la investigación, que se consideran como guías transversales del currículo. El propósito es proponer un marco de referencia que ilumine, contextualice y clarifique el sentido global de la enseñanza de la naturaleza de la ciencia a la luz de los contenidos de los currículos de ciencias españoles actuales. Primero, se presentan las contribuciones de algunos currículos extranjeros para identificar los contenidos consensuados que plantean. Segundo, se presenta la perspectiva y aportaciones de la competencia científica, desde los marcos de referencia de PISA y la competencia sobre conocimiento e interacción con el mundo físico. Tercero, se identifican los contenidos para la enseñanza de la naturaleza de la ciencia (y tecnología presentes en el currículo español de la educación secundaria y el bachillerato y se analiza su diseño, organización y sentido en el marco curricular español. Finalmente, se discuten las fortalezas y debilidades del marco curricular y se diagnostican las dificultades de la didáctica de estas cuestiones innovadoras sobre naturaleza de la ciencia (y tecnología, de modo que se pongan las bases que motiven, conciencien y ayuden al profesorado a superar las dificultades.

Appendix S-NH-1 and S-NH-2 of the experiment operating specification for the semiscale MOD-2C small break LOCA without HPI experiment series

International Nuclear Information System (INIS)

Owca, W.A.

1985-10-01

This document is Appendix S-NH--1 and S-NH-2 of the Experiment Operating Specification (EOS) for the Small Break LOCA without high pressure injection (HPI) series. It contains detailed information on the S-NH-1 and S-NH-2 experiment operation and facility configuration necessary to meet the series objectives stated in the main EOS body. 14 refs., 17 figs

Aproximaciones teóricas a la relación naturaleza - cultura: algunas contribuciones entre 1980 – 2004

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Ricardo Gómez

2009-10-01

Full Text Available El objetivo del presente artículo es mostrar algunas de las críticas más importantes dadas en el debate actual entre naturaleza y cultura. Busca reivindicar nuevos paradigmas de las ciencias dentro de la problemática del agotamiento de los recursos naturales del hogar que habitamos, la tierra, y el papel que juegan las comunidades locales contra el discurso predador del capitalismo global, a partir de su existencia cultural y social. Este trabajo se divide en tres partes que proponen por analogía, un símil entre los momentos de cambio de la noche al amanecer de un nuevo día; de la misma forma en que podríamos pasar de paradigmas “clásicos”, a procesos sociales recientes que incluyen los conocimientos locales. La primera, llamada “la noche”, expone brevísimamente los planteamientos más comunes que el capitalismo y las corrientes positivistas de la ciencia, mantienen frente a la naturaleza. La segunda, “el amanecer”, es una de las partes más importantes del presente escrito, ya que realiza una crítica a la oscuridad de la noche, a los planteamientos del capitalismo global desde el marxismo, el post-estructuralismo y el género. La tercera, “nuevo día”, concluye que a la luz de estas nuevas perspectivas de análisis podemos comprender mejor la relación entre los seres humanos y la tierra. Estos nuevos conocimientos políticos, culturales, ambientales y globales, desde el centro y la periferia, están atravesados por vínculos de poder que proponen una posición integradora entre conceptos aparentemente contradictorios y suponen susuperación a partir de la conciencia que autogeneran buscando orientar el tema político hacia estos nuevos conocimientos dentro de una agenda global.

Fast instrumentation for loss of coolant accident (LOCA) experimental studies pertaining to nuclear reactors

International Nuclear Information System (INIS)

Venkat Raj, V.; Sreenivas Rao, G.; Belokar, D.G.; Dolas, P.K.

1989-01-01

The loss of coolant accident (LOCA) which involves a breach in the pressure boundary of the primary coolant system (PCS) is one of the postulated accident conditions against which the safety of the reactor system is to be ensured. Mathematical models have been developed to analyse this kind of transients. However, because of the extremely complicated nature of the phenomena involved, it is necessary to validate the analytical models with appropriate experimental data. Many parameters are to be measured during the experiments, out of which temperature, pressure, void fraction and two-phase mass flow rate are the most important parameters. Since the phenomenon is very fast, special fast response instruments are required. This paper deals with the considerations that govern the selection of appropriate instruments and the development of suitable instruments for transient two-phase flow and void fraction measurements. The requirements of the associated fast data acquisition system are also discussed. (author). 4 figs

Large break LOCA analysis for retrofitted ECCS at MAPS using modified computer code ATMIKA

International Nuclear Information System (INIS)

Singhal, Mukesh; Khan, T.A.; Yadav, S.K.; Pramod, P.; Rammohan, H.P.; Bajaj, S.S.

2002-01-01

Full text: Computer code ATMIKA which has been used for thermal hydraulic analysis is based on unequal velocity equal temperature (UVET) model. Thermal hydraulic transient was predicted using three conservation equations and drift flux model. The modified drift flux model is now able to predict counter current flow and the relative velocity in vertical channel more accurately. Apart from this, stratification model is also introduced to predict the fuel behaviour under stratified condition. Many more improvements were carried out with respect to solution of conservation equation, heat transfer package and frictional pressure drop model. All these modifications have been well validated with published data on RD-12/RD-14 experiments. This paper describes the code modifications and also deals with the application of the code for the large break LOCA analysis for retrofitted emergency core cooling system (ECCS) being implemented at Madras Atomic Power Station (MAPS). This paper also brings out the effect of accumulator on stratification and fuel behaviour

ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

2010-02-01

A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

Large LOCA accident analysis for AP1000 under earthquake

International Nuclear Information System (INIS)

Yu, Yu; Lv, Xuefeng; Niu, Fenglei

2015-01-01

Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

Fuel Behaviour at High During RIA and LOCA Accidents

International Nuclear Information System (INIS)

Barrio del Juanes, M. T.; Garcia Cuesta, J. C.; Vallejo Diaz, I.; Herranz Puebla

2001-01-01

Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs

Evaluation of the pressure difference across the core during PWR-LOCA reflood phase

International Nuclear Information System (INIS)

Iguchi, Tadashi; Murao, Yoshio

1979-03-01

The flooding rate of the core influences largely cooling of the core during the reflood phase of a PWR-LOCA. Since the void fraction of two-phase flow in the core is important determining the flooding rate, it is essential to examine this void fraction. The void fraction in the core during the reflood phase obtained by experiment was compared with those predicted by the correlations respectively of Akagawa, Nicklin, Zuber, Yeh, Griffice, Behringer and Jhonson. Only Yeh's correlation was found to be usable for the purpose. The pressure difference of the core during the reflood phase was calculated by reflood analyzing code REFLA-1D using Yeh's correlation. Following are the results: (1) During the steady-state period after quenching of the heaters, the prediction agrees within +-15% with the experiment. (2) During the transient period when the quench front is advancing, the prediction is not in agreement with the experiment, the difference being about +-40%. Influence of the advancing quench front upon the void fraction in the core must further be studied. (author)

An analysis of LOCA sequences in the development of severe accident analysis DB

International Nuclear Information System (INIS)

Choi, Young; Park, Soo Yong; Ahn, Kwang-Il; Kim, D.H.

2006-01-01

Although a Level 2 PSA was performed for the Korean Standard Power Plants (KSNPs), and it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results. At present, KAERI is calculating the severe accident sequences intensively for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviours. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by knowledge-base technique, and the expected plant behaviour. The plant model used in this paper is oriented to the case of LOCAs related severe accident phenomena and thus can simulate the plant behaviours for a severe accident. Therefore the developed system may play a central role as an information source for decision-making for a severe accident management, and will be used as a training simulator for a severe accident management. (author)

Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

International Nuclear Information System (INIS)

Cheverton, R.D.

1976-09-01

The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall

The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

International Nuclear Information System (INIS)

Abdul-Razzak, A.; Zhang, J.; Sills, H.E.; Flatt, L.; Jenkins, D.; Wallace, D.J.; Popov, N.

2002-01-01

The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its proto-typing application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX R fuel. The methodology is consistent with and builds on world practice. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions. (authors)

Debris transport evaluation during the blow-down phase of a LOCA using computational fluid dynamics

International Nuclear Information System (INIS)

Park, Jong Pil; Jeong, Ji Hwan; Kim, Won Tae; Kim, Man Woong; Park, Ju Yeop

2011-01-01

Highlights: → We conducted CFD simulation on the spreading of the coolant in the containment after a break of the hot leg. It is used to estimate the dispersion of the debris within the containment. → It was assumed that the small and fine debris is transported by the discharge flow so that a fraction of the small and fine debris transport can be estimated based on the amount of water. → The break flow was assumed to be a homogeneous two-phase mixture without phase separation. Isenthalpic expansion of the break flow was used to specify the inlet boundary condition of the break flow. → The fraction of the small and fine debris transported to the upper part is 73%; this value is close to the value calculated using 1D lumped-parameter codes by the USNRC and the KINS, respectively, while 48% more than the value shown in the NEI 04-07. - Abstract: The performance of the emergency recirculation water sump under the influence of debris accumulation following a loss-of-coolant accident (LOCA) has long been of safety concern. Debris generation and transport during a LOCA are significantly influenced by the characteristics of the ejected coolant flow. One-dimensional analyses previously have been attempted to evaluate the debris transport during the blow-down phase but the transport evaluation still has large uncertainties. In this work, a computational fluid dynamics (CFD) analysis was utilized to evaluate small and fine debris transport during the blow-down phase of a pressurized water reactor, OPR1000. The coolant ejected from the ruptured hot-leg was assumed to expand in an isenthalpic process. The transport of small and fine debris was assumed to be dominated by water-borne transport, and the transport fractions for the upper and lower parts of the containment were quantified based on the CFD analysis. It was estimated that 73% of small and fine debris is transported to the upper part of the containment. This value is close to the values estimated by nuclear

La naturaleza como sujeto de derechos: análisis bioético de las Constituciones de Ecuador y Bolivia.

Directory of Open Access Journals (Sweden)

Irene Zasimowicz Pinto Calaça

2018-01-01

Full Text Available Las filosofías de la Pachamama y del buen vivir son puntos de referencia de los pueblos indígenas y están presentes en las constituciones de Ecuador y Bolivia, con el fin de armonizar las diferencias culturales y añadir las tradiciones a la política local. Las filosofías toman por base un mito andino: el que cree que la naturaleza es un organismo vivo y sujeto de derechos, y se abren espacio a la visión biocéntrica del mundo, compartida por la bioética global. La naturaleza ya no logra recomponerse por las innovaciones biotecnológicas impuestas por el hombre, lo que obliga a la humanidad a encontrar nuevos paradigmas, el Nuevo Constitucionalismo Latinoamericano, representado por las constituciones, es uno de ellos. En este artículo se describen y analizan críticamente estas dos constituciones, relacionándolas con la visión de mundo de la Pachamama y de la filosofía del buen vivir en el contexto de la bioética global. En este sentido, la Declaración Universal sobre Bioética y Derechos Humanos, al asociar bioética como campo normativo particular en la atención y cuidado de la vida y de la salud, con derechos humanos como campo normativo universal básico de obligaciones morales y jurídicas para todas las formas de vivir humano, presenta los valores fundamentales de una ética universal sustentada por la dignidad humana, la igualdad de derechos, la libertad, la justicia, la fraternidad y la paz, como es defendido en el artículo 17, de la protección del medio ambiente, de la biósfera y de la biodiversidad.

Reentrainment of droplet from grid spacer in mist flow portion of LOCA reflood of PWR

International Nuclear Information System (INIS)

Lee, S.L.; Cho, S.K.; Sheen, H.J.

1983-01-01

An investigation is made on the influence of a quenched grid spacer on the greatly enhanced heat transfer from heated fuel rods during the mist flow phase of emergency reflood of loss of coolant accident (LOCA) of a pressurized water reactor (PWR). The situation for the case of a dry grid spacer before its quenching has not been covered in this study. The experimental technique used is a relatively simple optical scheme which combines the reference-mode laser-Doppler anemometry making use of the scattering of a light beam from a droplet. The results reveal that the large droplets in the mist flow, which are intercepted by the grid spacer, are responsible for the creation of a large number of smaller droplets. These small droplets, due to their large surface area to mass ratios, can serve as superb evaporative cooling agents to heat transfer downstream of the grid spacer

Conservación de la naturaleza en propiedad privada: las Reservas Naturales de la Sociedad Civil en el Valle del Cauca

OpenAIRE

Melissa Quintero López; Fabio Alberto Arias Arbeláez

2016-01-01

Las Reservas Naturales de la Sociedad Civil en Colombia cumplen un papel importante en la conservación de ecosistemas. En este trabajo se investiga por qué los dueños de áreas privadas en el departamento del Valle del Cauca conservan la naturaleza en sus propiedades. Se consideran las explicaciones entre la teoría económica de la elección racional, la teoría económica sobre los comportamientos altruistas y recíprocos y la teoría de la psicología social sobre los valores de orientación cultura...

Significado estratigráfico y tectónico de los complejos de bloques resedimentados cambro-ordovicios de la Precordillera Occidental, Argentina

Directory of Open Access Journals (Sweden)

Vaccari, N. E.

1992-12-01

Full Text Available Rocks belonging to the Early Paleozoic Los Sombreros Formation crop out on the eastern flank of the Sierra de Yerba Loca, along the Río Jáchal section. This Formation consists of black shales and turbidites with interbedded limestone boulders near the base of the section, one of them up to 200 m long and 30 m thick. Boulders yield both early Cambrian trilobites and early Ordovician conodonts. Near the top, a thick sequence of dark grey marls and limestone contains an early Middle Cambrian trilobite fauna which is overlain by limestone and calcareous turbidites containing early Llanvirn graptolites and conodonts. On the basis of paleontological and sedimentological evidences this sequence can be interpreted as a series of submarine gravity slides interbedded within a dominantly turbiditic sequence. The Cambrian-early Ordovician precordilleran carbonate platform served as the source for much of the sediment which was redeposited on the slope. Then, deposition of the early Ordovician Los Sombreros Formation could be related to uplift and erosion of the platform margin and steeping of the slope during the late Arenig-early Llanvirn.En el flanco oriental de la Sierra de Yerba Loca, sobre el corte del río Jáchal, aflora una secuencia de edad paleozoica temprana correspondiente a los afloramientos más septentrionales de la Formación Los Sombreros. Esta unidad se compone de lutitas negras y turbiditas que contienen cerca de su base y hacia el techo horizontes de bloques calcáreos, algunos de enormes dimensiones -cerca de 30 m de espesor por más de 200 m de longitud- cuya fauna de trilobites y conodontos indica una edad Cámbrica y Ordovícica temprana. Hacia el techo de la sección está expuesta una potente secuencia de calcipelitas y calizas oscuras que contienen una fauna de trilobites del Cámbrico Medio basal, que es cubierta por brechas calcáreas y calciturbiditas portadoras de conodontos y graptolitos de edad llanvirniana temprana. Esta

Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

International Nuclear Information System (INIS)

Jiang, J.T.; Luxat, J.C.

2008-01-01

Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Jiang, J.T.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

2008-07-01

Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

International Nuclear Information System (INIS)

Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

2008-01-01

The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

Desenvolvimento e avaliação da eficácia de filmes biodegradáveis de amido de mandioca com nanocelulose como reforço e com extrato de erva-mate como aditivo antioxidante Development and evaluation of the effectiveness of biodegradable films of cassava starch with nanocelulose as reinforcement and yerba mate extract as an additive antioxidant

Directory of Open Access Journals (Sweden)

Bruna Aparecida Souza Machado

2012-11-01

Full Text Available O objetivo do trabalho foi desenvolver uma embalagem biodegradável utilizando como matriz polimérica o amido de mandioca plastificada com glicerol e reforçada com a incorporação de nanocelulose da fibra de coco, bem como, avaliar o efeito da adição de um aditivo natural (erva-mate nas formulações de nanobiocompósitos com ação antioxidante. Os nanocristais de celulose (L/D=39 foram obtidos por hidrólise ácida com H2SO4 a 65%. Os filmes foram preparados por casting contendo 4,5 e 6,0% de amido, 0,5 e 1,5% de glicerol, 0,3% de nanocelulose e 20% de extrato de erva-mate. O armazenamento do azeite de dendê embalado com os filmes contendo o aditivo foi monitorado por 40 dias sob condições de oxidação acelerada (63%UR/30°C. Constatou-se que, à medida que aumentam as perdas de Polifenóis Totais nos filmes, ocorre um menor aumento do Índice de Peróxidos do produto embalado, demonstrando, assim, que, ao invés do produto, os compostos da embalagem é quem estão sofrendo oxidação. A incorporação de extrato de erva-mate não alterou as propriedades mecânicas e de barreira desses filmes.The objective was to develop biodegradable packaging using a polymer matrix as the cassava starch plasticized with glycerol and reinforced with the incorporation of nanocelulose of coconut fiber, as well as to evaluate the effect of the addition of an additive nature (yerba mate in nanobiocompósitos formulations with antioxidant action. The nanocrystal cellulose (L/D=39 were obtained by acid hydrolysis with 65% H2SO4. The films were prepared by casting containing 4.5 and 6.0% starch, 0.5 and 1.5% glycerol, 0.3% nanocelulose and 20% extract of yerba mate. The palm oil storage packed with films containing the additive was monitored for 40 days under conditions of accelerated oxidation (63%UR/30°C. It was found that as the losses increase polyphenol films, there is a smaller increase of the peroxide value of the packaged product, thus

UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

International Nuclear Information System (INIS)

1998-01-01

1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

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Lázaro Lima

2011-12-01

Full Text Available

“Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

An assessment of post-LOCA radiolytic generation of hydrogen in reactor containment of Indian PHWRs

International Nuclear Information System (INIS)

Bose, H.; Shah, G.C.; Dutta, S.

2002-01-01

Full text: An event-wise assessment has been carried out for the 220 MWe Indian PHWRs of standardized design, to estimate the post-LOCA release of radiolytic hydrogen inside reactor containment, in absence of steam-zirconium reaction. The assessment is based on (i) the dissolved hydrogen concentration build-up in water corresponding to the decaying gamma dose profile and (ii) the rate of concentration dependent mass-transfer of hydrogen from water to gas-space. It is observed that the total radiolytic hydrogen released is about three times less than that obtained by the conventional method of calculation which assumes the radiolytic yield of hydrogen to be equal to the primary yield G(H 2 ) = 0.44 molecules per 100 eV. It is also seen that a major part (∼90 %) of the total release is due to the spillage of fission product irradiated suppression pool water flowing through the core, followed by moderator and suppression pool surface releases respectively

Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

International Nuclear Information System (INIS)

Nagler, A.; Gilat, J.; Hirshfeld, H.

1991-01-01

The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

Verification of LOCA/ECCS analysis codes ALARM-B2 and THYDE-B1 by comparison with RELAP4/MOD6/U4/J3

International Nuclear Information System (INIS)

Shimizu, Takashi

1982-08-01

For a verification study of ALARM-B2 code and THYDE-B1 code which are the component of the JAERI code system for evaluation of BWR ECCS performance, calculations for typical small and large break LOCA in BWR were done, and compared with those by RELAP4/MOD6/U4/J3 code. This report describes the influences of differences between the analytical models incorporated in the individual code and the problems identified by this verification study. (author)

Modelo sistémico para la conformación de un cluster turístico regional de naturaleza sustentable

Directory of Open Access Journals (Sweden)

Oscar Montaño-Arango

2012-01-01

Full Text Available Se presenta un modelo sistémico que visualiza y analiza los diferentes niveles y elementos que intervienen en el turismo de naturaleza sustentable como detonante del desarrollo regional, a través de un diagnóstico exploratorio, las vertientes que lo sustentan y la identificación de indicadores de potencialidad turística. Se detectan los actores que deben intervenir; así como la integración que debe tener el sector turístico, donde a través de la conceptualización, entendimiento y análisis del sistema, se generan alternativas para el desarrollo turístico regional con énfasis en los beneficios económicos, sociales, culturales y ambientales.