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Navegação Eventos - Artigos por assunto "accident-tolerant nuclear fuels"
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GOMES, DANIEL de S.
; GIOVEDI, CLAUDIA.
Assessment of advanced ferritic alloys used as cladding materials in nuclear power reactors.
In: INTERNATIONAL CONGRESS OF MECHANICAL ENGINEERING, 26th,
November 22-26, 2021,
Online.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Engenharia e Ciências Mecânicas - ABCM,
2021.
Abstract:
The fuel performance code, Fuel Analysis under Steady-state and Transients (FAST), permits cladding options, such as zirconium alloys and iron-chromium-aluminum (FeCrAl). FAST code support as cladding Kanthal, CM35, and CM36 alloys. We implemented a comparative analysis between ferritic alloys, steel, and zircaloy. Many features of ferritic alloys classify as more tolerant materials, such as high resistance to steam oxidation, reduced hydrogen release, and longer coping time. But the neutron penalty must reduce cladding thickness to let a greater fuel volume. Both ferritic alloys and austenitic steel show higher corrosion resistance, also avoiding hydrogen releases. FeCrAl provides more resistant corrosion cracking than stainless steel. The properties of steel 348 are comparable to those of FeCrAl alloys. Steel exhibits superior thermal conductivity, linear thermal expansion, and mechanical strength. Both offer similar specific heat, melting points, and densities. The chemical composition of the steel has 66% iron and 19% chromium, compared with Kanthal APMT™, which uses 68.8% iron and 22% chromium. The results found real advantages related to safety risks using ferritic cladding materials.
Palavras-Chave:
fuels;
stainless steels;
kanthal;
cladding;
accident-tolerant nuclear fuels;
zircaloy
GOMES, DANIEL de S.; GIOVEDI, CLAUDIA.
Assessment of advanced ferritic alloys used as cladding materials in nuclear power reactors.
In:
INTERNATIONAL CONGRESS OF MECHANICAL ENGINEERING, 26th,
November 22-26, 2021,
Online.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Engenharia e Ciências Mecânicas - ABCM,
2021.
Disponível em: http://repositorio.ipen.br/handle/123456789/32912. Acesso em: $DATA.
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GOMES, D.S.
; ABE, A.
; SILVA, A.T.
; MUNIZ, R.O.R.
; GIOVEDI, C.; MARTINS, M.R..
Assessment of high conductivity ceramic fuel concept under normal and accident conditions.
In: TECHNICAL MEETING ON MODELLING OF FUEL BEHAVIOUR IN DESIGN BASIS ACCIDENTS AND DESIGN EXTENSION CONDITIONS,
May 13-16, 2019,
Shenzhen, China.
Proceedings...
Vienna, Austria: International Atomic Energy Agency,
2020.
p. 95-101.
(IAEA-TECDOC-1913).
Abstract:
After the Fukushima Daiichi accident, the high conductivity ceramic concept fuel has been revisited. The thermal conductivity of uranium dioxide used as nuclear fuel is relatively low, as consequence fuel pellet centerline reaches high temperatures, high fission gas release rate, increase of fuel rod internal pressure reducing the safety thermal margin. Several investigations had been conducted in framework of ATF (Accident Tolerant Fuel) using different additives in ceramic fuel (UO2) in order to enhance thermal conductivity in uranium dioxide pellets. The increase of the thermal conductivity of fuel can reduce the pellet centerline temperature, consequently less fission gas releasing rate and the low risk of fuel melting, hence improving significantly fuel performance under accident conditions. The beryllium oxide (BeO) has high conductivity among other ceramics and is quite compatible with UO2up to 2200°C, at which temperature it forms a eutectic. Moreover, it is compatible with zircaloy cladding, does not react with water, has a good neutronic characteristics (low neutron absorption cross-section, neutron moderation). This work presents a preliminary assessment of high conductivity ceramic concept fuel considering UO2-BeO mixed oxide fuel containing 10 wt% of BeO. The FRAPCON and FRAPTRAN fuel performance codes were conveniently adapted to support the evaluation of UO2-BeO mixed oxide fuel. The thermal and mechanical properties were modified in the codes for a proper and representative simulation of the fuel performance. Theobtainedpreliminary results show lower fuel centerline temperatureswhen compared to standard UO2 fuel, consequently promoting enhancement of safety margins during the operational condition and under LOCA accident scenario.
Palavras-Chave:
accident-tolerant nuclear fuels;
beryllium oxides;
uranium dioxide;
zircaloy;
ceramics;
cladding;
cross sections;
eutectics;
fission product release;
fission products;
fuel cans;
fuel pellets;
fuel rods;
fukushima daiichi nuclear power station;
loss of coolant;
mechanical properties;
melting;
mixed oxide fuels;
performance;
safety margins;
simulation;
thermal conductivity
GOMES, D.S.; ABE, A.; SILVA, A.T.; MUNIZ, R.O.R.; GIOVEDI, C.; MARTINS, M.R.
Assessment of high conductivity ceramic fuel concept under normal and accident conditions.
In:
TECHNICAL MEETING ON MODELLING OF FUEL BEHAVIOUR IN DESIGN BASIS ACCIDENTS AND DESIGN EXTENSION CONDITIONS,
May 13-16, 2019,
Shenzhen, China.
Proceedings...
Vienna, Austria: International Atomic Energy Agency,
2020.
p. 95-101.
(IAEA-TECDOC-1913).
Disponível em: http://repositorio.ipen.br/handle/123456789/31064. Acesso em: $DATA.
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Como referenciar este item
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GIOVEDI, C.; MARTINS, M.R.; ABE, A.
; MUNIZ, R.O.R.
; GOMES, D.S.
; SILVA, A.T.
.
Fuel performance assessment of enhanced accident tolerant fuel using iron-based alloys as cladding.
In: TOPFUEL,
30 September - 04 October, 2018,
Prague, Czech Republic.
Proceedings...
Brussels, Belgium: European Nuclear Society,
2018.
Abstract:
In the framework of the Enhanced Accident Tolerant Fuel (EATF) program, one
important tool to assess the behaviour of new materials under irradiation is the use
of fuel performance codes. For this, it is necessary to modify conventional fuel
performance codes to introduce the properties of the materials to be studied. The
aim of this paper is to present some preliminary results obtained using modified
versions of the FRAPCON code adapted to evaluate the performance as cladding
of two different types of iron-based alloys as cladding: stainless steel (AISI 348),
and FeCrAl alloy, including a preliminary sensitivity analysis. The results obtained
using the modified versions of the codes were compared to those obtained for
zirconium-based alloys using the original code version. The results have shown
and confirmed that iron-based alloys are one of the promising candidates to be
used as EATF cladding in PWR.
Palavras-Chave:
iron base alloys;
accident-tolerant nuclear fuels;
fuel-cladding interactions
GIOVEDI, C.; MARTINS, M.R.; ABE, A.; MUNIZ, R.O.R.; GOMES, D.S.; SILVA, A.T.
Fuel performance assessment of enhanced accident tolerant fuel using iron-based alloys as cladding.
In:
TOPFUEL,
30 September - 04 October, 2018,
Prague, Czech Republic.
Proceedings...
Brussels, Belgium: European Nuclear Society,
2018.
Disponível em: http://repositorio.ipen.br/handle/123456789/31126. Acesso em: $DATA.
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GOMES, DANIEL S.
; MUNIZ, RAFAEL O.R.
; GIOVEDI, CLAUDIA.
Improving performance with accident tolerant-fuels.
In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Abstract:
After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO2–Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density—above that supported by UO2—and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U3Si2, UN, and UC contain a higher uranium density and thermal conductivity than UO2, providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U3Si2, UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO2. These ATFs also have thermal conductivities approximately four times higher than that of UO2. A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO2–Zr as the fuel system of choice.
Palavras-Chave:
accident-tolerant nuclear fuels;
aluminium;
chromium;
cladding;
comparative evaluations;
iron;
oxidation;
physical properties;
swelling;
uranium oxides;
water cooled reactors;
zirconium alloys
GOMES, DANIEL S.; MUNIZ, RAFAEL O.R.; GIOVEDI, CLAUDIA.
Improving performance with accident tolerant-fuels.
In:
INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Disponível em: http://repositorio.ipen.br/handle/123456789/28190. Acesso em: $DATA.
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GIOVEDI, CLAUDIA; ABE, ALFREDO
; MUNIZ, RAFAEL O.R.
; GOMES, DANIEL de S.
; SILVA, ANTONIO T. e
; MARTINS, MARCELO R..
Modification of fuel performance code to evaluate iron-based alloy behavior under loca scenario.
In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Abstract:
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 ironbased
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.
Palavras-Chave:
accident-tolerant nuclear fuels;
cladding;
computerized simulation;
f codes;
fuel rods;
iron alloys;
loss of coolant;
performance;
pwr type reactors;
stainless steel-348
GIOVEDI, CLAUDIA; ABE, ALFREDO; MUNIZ, RAFAEL O.R.; GOMES, DANIEL de S.; SILVA, ANTONIO T. e; MARTINS, MARCELO R.
Modification of fuel performance code to evaluate iron-based alloy behavior under loca scenario.
In:
INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Disponível em: http://repositorio.ipen.br/handle/123456789/28187. Acesso em: $DATA.
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-
ABE, ALFREDO Y.
; MELO, CAIO; GIOVEDI, CLAUDIA; SILVA, ANTONIO T.
.
Modification of TRANSURANUS fuel performance code in the ATF framework.
In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 21-25, 2019,
Santos, SP.
Proceedings...
Rio de Janeiro: Associação Brasileira de Energia Nuclear,
2019.
p. 5036-5045.
Abstract:
The standard fuel system based on UO2–zirconium alloy has been utilized on nearly 90% of worldwide nuclear power light water reactors. After the Fukushima Daiichi accident, alternative cladding materials to zirconium-based alloys are being investigated in the framework of accident tolerance fuel (ATF) program. One of the concepts of ATF is related to cladding materials that could delay the onset of high temperature oxidation, as well as ballooning and burst, in order to improve reactor safety systems, and consequently increase the coping time for the reactor operators in accident condition, especially under Loss-of-Coolant Accident (LOCA) scenario. The ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium-based alloys based on its outstanding resistance to oxidation under superheated steam environment due to the development of alumina oxide on the alloy surface in case of LOCA; moreover, FeCrAl alloys present quite well performance under normal operation conditions due to the thin oxide rich in chromium that acts as a protective layer. The assessment and performance of new fuel systems rely on experimental irradiation program and fuel performance code simulation, therefore the aim of this work is to contribute to the computational modeling capabilities in the framework of the ATF concept. The well-known TRANSURANUS fuel performance code that is used by safety authorities, industries, laboratories, research centers and universities was modified in order to support FeCrAl alloy as cladding material. The modification of the TRANSURANUS code was based on existing data (material properties) from open literature and as verification process was performed considering LOCA accident scenario.
Palavras-Chave:
accident-tolerant nuclear fuels;
aluminium alloys;
chromium alloys;
cladding;
comparative evaluations;
fuel rods;
iron alloys;
loss of coolant;
performance;
t codes;
zirconium alloys
ABE, ALFREDO Y.; MELO, CAIO; GIOVEDI, CLAUDIA; SILVA, ANTONIO T.
Modification of TRANSURANUS fuel performance code in the ATF framework.
In:
INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 21-25, 2019,
Santos, SP.
Proceedings...
Rio de Janeiro: Associação Brasileira de Energia Nuclear,
2019.
p. 5036-5045.
Disponível em: http://repositorio.ipen.br/handle/123456789/30814. Acesso em: $DATA.
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-
GIOVEDI, C.; MARTINS, M.R.; ABE, A.
; REIS, R.
; SILVA, A.T.
.
Reactivity initiated accident assessment for ATF cladding materials.
In: TECHNICAL MEETING ON MODELLING OF FUEL BEHAVIOUR IN DESIGN BASIS ACCIDENTS AND DESIGN EXTENSION CONDITIONS,
May 13-16, 2019,
Shenzhen, China.
Proceedings...
Vienna, Austria: International Atomic Energy Agency,
2020.
p. 155-161.
(IAEA-TECDOC-1913).
Abstract:
Following the experience that came from the Fukushima Daiichi accident, one possible way of reducing risk in a nuclear power plant operation would be the replacement of the existing fuel rod cladding material (based on zirconium alloys) by another materials which could fulfill the requirements of the accident tolerant fuel (ATF) concept. In this sense, ATF should be able to keep the current fuel system performance under normal operation conditions; moreover, it should present superior performance than the existing conventional fuel system (zirconium-based alloys and uranium dioxide) under accident conditions. The most challenging and bounding accident scenarios for nuclear fuel systems in Pressurized Water Reactors (PWR) are Loss of Coolant Accident (LOCA) and Reactivity Initiated Accident (RIA), which are postulated accidents. This work addresses the performance of ATF using iron-based alloys as cladding material under RIA conditions. The evaluation is carried out using modified versions of the coupled system FRAPCON/FRAPTRAN. These codes were modified to include the material properties (thermal, mechanical, and physics) of an iron-based alloy, specifically FeCrAl alloy. The analysis is performed using data available in the open literature related to experiments using conventional PWR fuel system (zirconium-based alloys and uranium dioxide). The results obtained using the modified code versions are compared to those of the actual existing fuel system based on zircaloy-4 cladding using the original versions of the fuel performance codes (FRAPCON/FRAPTRAN).
Palavras-Chave:
accident-tolerant nuclear fuels;
charges;
cladding;
comparative evaluations;
currents;
fuel rods;
fuel systems;
fukushima daiichi nuclear power station;
hazards;
zircaloy 4;
iron;
loss of coolant;
nuclear power plants;
operation;
performance;
pwr type reactors;
reactivity-initiated accidents;
steady-state conditions;
uranium dioxide
GIOVEDI, C.; MARTINS, M.R.; ABE, A.; REIS, R.; SILVA, A.T.
Reactivity initiated accident assessment for ATF cladding materials.
In:
TECHNICAL MEETING ON MODELLING OF FUEL BEHAVIOUR IN DESIGN BASIS ACCIDENTS AND DESIGN EXTENSION CONDITIONS,
May 13-16, 2019,
Shenzhen, China.
Proceedings...
Vienna, Austria: International Atomic Energy Agency,
2020.
p. 155-161.
(IAEA-TECDOC-1913).
Disponível em: http://repositorio.ipen.br/handle/123456789/31065. Acesso em: $DATA.
Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.
Como referenciar este item
-
ABE, ALFREDO
; GIOVEDI, CLAUDIA; GOMES, DANIEL
; SILVA, ANTONIO T. e
; MUNIZ, RAFAEL O.R.
; MARTINS, MARCELO.
Sensitivity assessment of fuel performance codes for loca accident scenario.
In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Abstract:
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.
Palavras-Chave:
accident-tolerant nuclear fuels;
computerized simulation;
f codes;
fuel rods;
fuel-cladding interactions;
loss of coolant;
sensitivity analysis;
transients;
water cooled reactors
ABE, ALFREDO; GIOVEDI, CLAUDIA; GOMES, DANIEL; SILVA, ANTONIO T. e; MUNIZ, RAFAEL O.R.; MARTINS, MARCELO.
Sensitivity assessment of fuel performance codes for loca accident scenario.
In:
INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Disponível em: http://repositorio.ipen.br/handle/123456789/28196. Acesso em: $DATA.
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-
GOMES, DANIEL S.
; SILVA, ANTONIO T.
; ABE, ALFREDO Y.
; MUNIZ, RAFAEL O.R.
; GIOVEDI, CLAUDIA.
Simulation of accident-tolerant U3Si2 fuel using FRAPCON code.
In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Abstract:
The research on accident-tolerant fuels (ATFs) increased after the Fukushima event. This benefitted risk management in nuclear operations. In this investigation, the physical properties of the materials being developed for the ATF program were compared with those of the standard UO2–Zr fuel system. The research efforts in innovative fuel design include rigorous characterization of thermal, mechanical, and chemical assessment, with the objectives of making the burnup cycle longer, increasing power density, and improving safety performance. Fuels must reach a high uranium density—above that supported by UO2—and possess coating that exhibits better oxidation resistance than Zircaloy. The uranium density and thermal conductivity of ATFs, such as U3Si2, UN, and UC, is higher than that of UO2; their combination with advanced cladding provides possible fuel–cladding options. An ideal combination of fuel and cladding must increase fuel performance in loss-of-coolant scenarios. The disadvantages of U3Si2, UN, and UC are their swelling rates, which are higher than that of UO2. The thermal conductivities of ATFs are approximately four times higher than that of UO2. To prevent the generation of hydrogen due to oxidation of zirconium-based alloys in contact with steam, cladding options, such as ferritic alloys, were studied. It was verified that FeCrAl alloys and SiC provide better response under severe conditions because of their thermophysical properties. The findings of this study indicate that U3Si2 and the FeCrAl fuel cladding concept should replace UO2–Zr as the fuel system of choice.
Palavras-Chave:
accident-tolerant nuclear fuels;
aluminium alloys;
chromium alloys;
cladding;
comparative evaluations;
computerized simulation;
f codes;
fuel rods;
iron alloys;
loss of coolant;
steady-state conditions;
swelling;
thermal conductivity;
thermal expansion;
transients;
uranium carbides;
uranium nitrides;
uranium silicides;
zircaloy
GOMES, DANIEL S.; SILVA, ANTONIO T.; ABE, ALFREDO Y.; MUNIZ, RAFAEL O.R.; GIOVEDI, CLAUDIA.
Simulation of accident-tolerant U3Si2 fuel using FRAPCON code.
In:
INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE,
October 22-27, 2017,
Belo Horizonte, MG.
Proceedings...
Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear,
2017.
Disponível em: http://repositorio.ipen.br/handle/123456789/28188. Acesso em: $DATA.
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ATENÇÃO!
ESTE TEXTO "AJUDA" ESTÁ SUJEITO A ATUALIZAÇÕES CONSTANTES, A MEDIDA QUE NOVAS FUNCIONALIDADES E RECURSOS DE BUSCA FOREM SENDO DESENVOLVIDOS PELAS EQUIPES DA BIBLIOTECA E DA INFORMÁTICA.
O gerenciamento do Repositório está a cargo da Biblioteca do IPEN. Constam neste RI, até o presente momento 20.950 itens que tanto podem ser artigos de periódicos ou de eventos nacionais e internacionais, dissertações e teses, livros, capítulo de livros e relatórios técnicos. Para participar do RI-IPEN é necessário que pelo menos um dos autores tenha vínculo acadêmico ou funcional com o Instituto. Nesta primeira etapa de funcionamento do RI, a coleta das publicações é realizada periodicamente pela equipe da Biblioteca do IPEN, extraindo os dados das bases internacionais tais como a Web of Science, Scopus, INIS, SciElo além de verificar o Currículo Lattes. O RI-IPEN apresenta também um aspecto inovador no seu funcionamento. Por meio de metadados específicos ele está vinculado ao sistema de gerenciamento das atividades do Plano Diretor anual do IPEN (SIGEPI). Com o objetivo de fornecer dados numéricos para a elaboração dos indicadores da Produção Cientifica Institucional, disponibiliza uma tabela estatística registrando em tempo real a inserção de novos itens. Foi criado um metadado que contém um número único para cada integrante da comunidade científica do IPEN. Esse metadado se transformou em um filtro que ao ser acionado apresenta todos os trabalhos de um determinado autor independente das variáveis na forma de citação do seu nome.
A elaboração do projeto do RI do IPEN foi iniciado em novembro de 2013, colocado em operação interna em julho de 2014 e disponibilizado na Internet em junho de 2015. Utiliza o software livre Dspace, desenvolvido pelo Massachusetts Institute of Technology (MIT). Para descrição dos metadados adota o padrão Dublin Core. É compatível com o Protocolo de Arquivos Abertos (OAI) permitindo interoperabilidade com repositórios de âmbito nacional e internacional.
1. Portaria IPEN-CNEN/SP nº 387, que estabeleceu os princípios que nortearam a criação do RDI,
clique aqui.
2. A experiência do Instituto de Pesquisas Energéticas e Nucleares (IPEN-CNEN/SP) na criação de um Repositório Digital Institucional – RDI,
clique aqui.
O Repositório Digital do IPEN é um equipamento institucional de acesso aberto, criado com o objetivo de reunir, preservar, disponibilizar e conferir maior visibilidade à Produção Científica publicada pelo Instituto, desde sua criação em 1956.
Operando, inicialmente como uma base de dados referencial o Repositório foi disponibilizado na atual plataforma, em junho de 2015. No Repositório está disponível o acesso ao conteúdo digital de artigos de periódicos, eventos, nacionais e internacionais, livros, capítulos, dissertações, teses e relatórios técnicos.
A elaboração do projeto do RI do IPEN foi iniciado em novembro de 2013, colocado em operação interna em julho de 2014 e disponibilizado na Internet em junho de 2015. Utiliza o software livre Dspace, desenvolvido pelo Massachusetts Institute of Technology (MIT). Para descrição dos metadados adota o padrão Dublin Core. É compatível com o Protocolo de Arquivos Abertos (OAI) permitindo interoperabilidade com repositórios de âmbito nacional e internacional.
O gerenciamento do Repositório está a cargo da Biblioteca do IPEN. Constam neste RI, até o presente momento 20.950 itens que tanto podem ser artigos de periódicos ou de eventos nacionais e internacionais, dissertações e teses, livros, capítulo de livros e relatórios técnicos. Para participar do RI-IPEN é necessário que pelo menos um dos autores tenha vínculo acadêmico ou funcional com o Instituto. Nesta primeira etapa de funcionamento do RI, a coleta das publicações é realizada periodicamente pela equipe da Biblioteca do IPEN, extraindo os dados das bases internacionais tais como a Web of Science, Scopus, INIS, SciElo além de verificar o Currículo Lattes. O RI-IPEN apresenta também um aspecto inovador no seu funcionamento. Por meio de metadados específicos ele está vinculado ao sistema de gerenciamento das atividades do Plano Diretor anual do IPEN (SIGEPI). Com o objetivo de fornecer dados numéricos para a elaboração dos indicadores da Produção Cientifica Institucional, disponibiliza uma tabela estatística registrando em tempo real a inserção de novos itens. Foi criado um metadado que contém um número único para cada integrante da comunidade científica do IPEN. Esse metadado se transformou em um filtro que ao ser acionado apresenta todos os trabalhos de um determinado autor independente das variáveis na forma de citação do seu nome.