IO and OI. II

DEFF Research Database (Denmark)

Engelfriet, Joost; Schmidt, Erik Meineche

1978-01-01

In Part 1 of this paper (J. Comput. System Sci. 15, Number 3 (1977)) we presented a fixed point characterization of the (IO and OI) context-free tree languages. We showed that a context-free tree grammar can be viewed as a system of regular equations over a tree language substitution algebra. In ....... In this part we shall use these results to obtain a theory of systems of context-free equations over arbitrary continuous algebras. We refer to the Introduction of Part 1 for a description of the contents of this part.......In Part 1 of this paper (J. Comput. System Sci. 15, Number 3 (1977)) we presented a fixed point characterization of the (IO and OI) context-free tree languages. We showed that a context-free tree grammar can be viewed as a system of regular equations over a tree language substitution algebra...

The europium(II) oxide halides Eu{sub 2}OBr{sub 2} and Eu{sub 2}OI{sub 2}; Die Europium(II)-Oxidhalogenide Eu{sub 2}OBr{sub 2} und Eu{sub 2}OI{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Rudolph, Daniel; Schleid, Thomas [Stuttgart Univ. (Germany). Inst. fuer Anorganische Chemie

2017-07-01

The syntheses and crystal structures of the two isotypic europium(II) oxide halides Eu{sub 2}OBr{sub 2} and Eu{sub 2}OI{sub 2} are reported. They crystallize orthorhombically in the space group Ibam (Z=4; Eu{sub 2}OBr{sub 2}: a=709.86(5), b=1200.34(9), c=628.71(4) pm; Eu{sub 2}OI{sub 2}: a=739.78(5), b=1295.13(9), c=644.82(4) pm). The unit cell parameters presented here, and thus the interatomic distances of Eu{sub 2}OI{sub 2}, are significantly smaller than the ones reported in the literature, which is explained by the substitution of europium with larger barium cations due to the synthesis route described in the early study. Central building blocks of both crystal structures are trans-edge-connected [OEu{sub 4}]{sup 6+} tetrahedra forming straight {sup 1}{sub ∞}{[OEu"e_4_/_2]"2"+} chains running parallel to the [001] direction. Bundled like a hexagonal rod packing, their interaction is achieved by Br{sup -} or I{sup -} anions for charge compensation.

Myths about OI (Osteogenesis Imperfecta)

Science.gov (United States)

... Based on the OI Foundation publication Introduction to Osteogenesis Imperfecta: A Guide for Medical Professionals, Individuals and Families ... for Children, editor, 2013. Page updated August, 2015. © Osteogenesis Imperfecta Foundation, 2015 Privacy Policy

BiOI/TiO2-nanorod array heterojunction solar cell: Growth, charge transport kinetics and photoelectrochemical properties

International Nuclear Information System (INIS)

Wang, Lingyun; Daoud, Walid A.

2015-01-01

Highlights: • BiOI/TiO 2 photoanodes were fabricated by a simple solvothermal/hydrothermal method. • BiOI/TiO 2 (PVP) showed a 13-fold increase in photocurrent density compared to TiO 2 . • Charge transport kinetics within the BiOI/TiO 2 heterojunctions are discussed. - Abstract: A series of BiOI/TiO 2 -nanorod array photoanodes were grown on fluorine-doped tin oxide (FTO) glass using a simple two-step solvothermal/hydrothermal method. The effects of the hydrothermal process, such as TiO 2 nanorod growth time, BiOI concentration and the role of surfactant, polyvinylpyrrolidone (PVP), on the growth of BiOI, were investigated. The heterojunctions were characterized by X-ray diffraction, UV–vis absorbance spectroscopy and scanning electron microscopy. The photoelectrochemical properties of the as-grown junctions, such as linear sweep voltammetry (LSV) behavior, photocurrent response and incident photon-to-electron conversion efficiency (IPCE) under Xenon lamp illumination, are presented. The cell with BiOI/TiO 2 (PVP) as photoanode can reach a short current density (J sc ) of 0.13 mA/cm 2 and open circuit voltage (V oc ) of 0.46 V vs. Ag/AgCl under the irradiation of a 300 W Xenon lamp. Compared to bare TiO 2 , the IPCE of BiOI/TiO 2 (PVP) increased 4–5 times at 380 nm. Furthermore, the charge transport kinetics within the heterojunction is also discussed

In memory of Jean-François Stéphan

Science.gov (United States)

Blanchet, René

2016-01-01

This thematic issue of Comptes rendus Geoscience has been assembled to honor the memory of our late colleague and friend Jean-François Stéphan, whose remarkable scientific and community-directed activity has left a deep imprint on both the French and the International Earth Science communities. This volume brings together contributions of colleagues of Jean-François who were also close friends. Naturally, tectonics is the common theme of these contributions. Some of the papers presented here focus on tectonic questions and/or regions Jean-François worked on during his career; other papers present studies Jean-François motivated or encouraged in one way or another. Taken together, the papers of this thematic issue take the reader on a beautiful trip, from past to current tectonics.

Design and construction of earth retaining walls with anchors employed in excavation works at Oi Nuclear Power Plant

International Nuclear Information System (INIS)

Saino, Susumu; Aoshima, Ken-ichiro; Kamide, Atsushi.

1990-01-01

In Oi Nuclear Power Station, Kansai Electric Power Co., Inc., No.3 and No.4 plants of each 1180 MWe output are additionally installed, neighboring existing No.1 and No.2 plants of each 1175 MWe output in operation. The start of operation is expected in December, 1991 in No.3 plant, and in February, 1993 in No.4 plant. The total quantity of earth excavated for this additional installation works is about 3.3 million m 3 . The main works are, subsequently to the preparation of the site, the excavation for the foundations of reactor buildings and others, and the construction of the foundations for the seawater system facilities for cooling condensers and reactor auxiliary machines, and the works were begun in May, 1987. The excavation by using anchors was carried out in seven places. The vertical excavation on large scale was carried out by using the earth retaining walls of concrete-sprayed anchor structure in drain pits. In this report, the outline of the geological features, the outline of the excavation works, the design of the earth retaining walls, the execution of concrete spraying, the planning and result of measurement are described. (K.I.)

François Grabowski ( 1939–2008 )

CERN Multimedia

2008-01-01

We are saddened by the news that our colleague François Grabowski passed away at the end of April. François joined CERN in February 1968, and as a member of the RF Separators Group in TC Division he participated in the construction of the separator on the beam of the Mirabelle bubble chamber and its installation at Serpukhov (former Soviet Union). He also helped build the shielding for the West Area neutrino beam and the 6 GHz RF separator installed on the separated beam for the Big European Bubble Chamber (BEBC). During the 80s, as a member of the Superconducting Cavities Group of EF Division (subsequently to become the AT Department), he took part in the development and construction of the LEP200 cavities. Finally, as a member of the SL Division, he went on to work on the superconducting cavities for the LHC. François retired from CERN in July 2004 but was sadly unable to make the most of the time remaining to him as he battled ...

Heterojunction BiOI/Bi2MoO6 nanocomposite with much enhanced photocatalytic activity

International Nuclear Information System (INIS)

Li, Wen Ting; Zheng, Yi Fan; Yin, Hao Yong; Song, Xu Chun

2015-01-01

BiOI/Bi 2 MoO 6 heterostructures with different amounts of BiOI were successfully prepared via a facile deposition method. The obtained BiOI/Bi 2 MoO 6 photocatalysts exhibited much higher visible light (λ > 420 nm) induced photocatalytic activity compared with single Bi 2 MoO 6 and BiOI photocatalysts. 20 % BiOI/Bi 2 MoO 6 nanocomposite exhibited the highest photocatalytic activity with almost all RhB decomposed within 70 min. However, excess BiOI covering on the surface of Bi 2 MoO 6 can inversely reduce the photocatalytic activity. The enhanced photocatalytic activities could be resulted from the function of the novel p–n heterojunction interface between Bi 2 MoO 6 and BiOI, which could separate photoinduced carriers efficiently. Possible mechanisms on the basis of the relative band positions were also discussed

An investigation of the ionospheric F region near the EIA crest in India using OI 777.4 and 630.0 nm nightglow observations

Directory of Open Access Journals (Sweden)

N. Parihar

2018-05-01

Full Text Available Simultaneous observations of OI 777.4 and OI 630.0 nm nightglow emissions were carried at a low-latitude station, Allahabad (25.5° N, 81.9° E; geomag. lat.  ∼  16.30° N, located near the crest of the Appleton anomaly in India during September–December 2009. This report attempts to study the F region of ionosphere using airglow-derived parameters. Using an empirical approach put forward by Makela et al. (2001, firstly, we propose a novel technique to calibrate OI 777.4 and 630.0 nm emission intensities using Constellation Observing System for Meteorology, Ionosphere, and Climate/Formosa Satellite Mission 3 (COSMIC/FORMOSAT-3 electron density profiles. Next, the electron density maximum (Nm and its height (hmF2 of the F layer have been derived from the information of two calibrated intensities. Nocturnal variation of Nm showed the signatures of the retreat of the equatorial ionization anomaly (EIA and the midnight temperature maximum (MTM phenomenon that are usually observed in the equatorial and low-latitude ionosphere. Signatures of gravity waves with time periods in the range of 0.7–3.0 h were also seen in Nm and hmF2 variations. Sample Nm and hmF2 maps have also been generated to show the usefulness of this technique in studying ionospheric processes.

BiOI/TiO{sub 2}-nanorod array heterojunction solar cell: Growth, charge transport kinetics and photoelectrochemical properties

Energy Technology Data Exchange (ETDEWEB)

Wang, Lingyun; Daoud, Walid A., E-mail: wdaoud@cityu.edu.hk

2015-01-01

Highlights: • BiOI/TiO{sub 2} photoanodes were fabricated by a simple solvothermal/hydrothermal method. • BiOI/TiO{sub 2} (PVP) showed a 13-fold increase in photocurrent density compared to TiO{sub 2}. • Charge transport kinetics within the BiOI/TiO{sub 2} heterojunctions are discussed. - Abstract: A series of BiOI/TiO{sub 2}-nanorod array photoanodes were grown on fluorine-doped tin oxide (FTO) glass using a simple two-step solvothermal/hydrothermal method. The effects of the hydrothermal process, such as TiO{sub 2} nanorod growth time, BiOI concentration and the role of surfactant, polyvinylpyrrolidone (PVP), on the growth of BiOI, were investigated. The heterojunctions were characterized by X-ray diffraction, UV–vis absorbance spectroscopy and scanning electron microscopy. The photoelectrochemical properties of the as-grown junctions, such as linear sweep voltammetry (LSV) behavior, photocurrent response and incident photon-to-electron conversion efficiency (IPCE) under Xenon lamp illumination, are presented. The cell with BiOI/TiO{sub 2} (PVP) as photoanode can reach a short current density (J{sub sc}) of 0.13 mA/cm{sup 2} and open circuit voltage (V{sub oc}) of 0.46 V vs. Ag/AgCl under the irradiation of a 300 W Xenon lamp. Compared to bare TiO{sub 2}, the IPCE of BiOI/TiO{sub 2} (PVP) increased 4–5 times at 380 nm. Furthermore, the charge transport kinetics within the heterojunction is also discussed.

OI Issues: Type I - Understanding the Mildest Form of Osteogenesis Imperfecta

Science.gov (United States)

... Issues: Type I—Understanding the Mildest Form of Osteogenesis Imperfecta Type I OI Osteogenesis imperfecta (OI) is a ... 223-0344 Toll free: 800-624-BONE (2663) Osteogenesis Imperfecta Foundation Website: http://www.oif.org The National ...

Preparation of O/I1-type Emulsions and S/I1-type Dispersions Encapsulating UV-Absorbing Agents.

Science.gov (United States)

Aramaki, Kenji; Kimura, Minami; Masuda, Kazuki

2015-01-01

Oil-in-cubic phase (O/I1) emulsions encapsulating the cosmetic UV absorbing agents 2-ethylhexyl 4-methoxycinnamate (EHMC), 2-ethylhexyl 2-cyano-3,3-diphenylacrylate (octocrylene, OCR) and 1-(4-tertbutylphenyl)-3-(4-methoxyphenyl)-1,3-propanedione (Avobenzone, TBMP) were prepared by vortex mixing accompanied by a heating-cooling process. A ternary phase diagram in a water/C12EO25/EHMC system at 25°C was constructed and the two-phase equilibrium of an oil phase and an I1 phase, which is necessary to prepare the O/I1-type emulsions, was confirmed. Also, the melting of the I1 phase into a fluid micellar solution phase was confirmed, allowing emulsification by a heating-cooling process. The O/I1-type emulsions were formulated in the ternary system as well as a quaternary system. The four-component system contained an additional cosolvent, isopropyl myristate (IPM). The use of the cosolvent allows the use of reduced amounts of EHMC, which is desirable because EHMC can cause temporary skin irritation. Formulation of the O/I1-type emulsions with other UV absorbing agents (OCR and TBMP) was also possible using the same emulsification method. When IPM was changed to tripalmitin, which has a melting point greater than room temperature, a solid-oil dispersion in I1 phase was formed. We have termed this a "solidin-cubic phase (S/I1) type dispersion". These novel emulsions have not been reported previously. The UV absorbability of the O/I1-type emulsions and S/I1-type dispersions that encapsulate the UV absorbing agents was confirmed by measurement of UV absorption spectra.

François Morel (1923-2007)

OpenAIRE

Corvol, Pierre

2010-01-01

Le 9 mai 2007, François Morel nous quittait à l’âge de 84 ans, après toute une vie consacrée à la recherche sur la physiologie rénale qu’il a réalisée en grande partie au Collège de France où il a été titulaire de la chaire de Physiologie cellulaire de 1967 à 1993. François Morel était né à Genève en 1923. Son père était titulaire de la chaire de Psychiatrie dans cette ville, ce qui l’a sans doute incité à entreprendre des études de médecine. En fait, il n’a jamais exercé la médecine car, trè...

In-situ synthesis of nanofibers with various ratios of BiOCl{sub x}/BiOBr{sub y}/BiOI{sub z} for effective trichloroethylene photocatalytic degradation

Energy Technology Data Exchange (ETDEWEB)

Zhang, Yifan [Department of Chemistry, Inha University, 100 Inharo, Incheon 402-751 (Korea, Republic of); Park, Mira [Department of Organic Materials and Fiber Engineering, Chonbuk National University, Jeonju 561-756 (Korea, Republic of); Kim, Hak Yong [Department of BIN Convergence Technology, Chonbuk National University, Jeonju, 561-756 (Korea, Republic of); Ding, Bin [College of Textiles, Donghua University, Shanghai 201620 (China); Park, Soo-Jin, E-mail: sjpark@inha.ac.kr [Department of Chemistry, Inha University, 100 Inharo, Incheon 402-751 (Korea, Republic of)

2016-10-30

Highlights: • BiOCl{sub x}/BiOBr{sub y}/BiOI{sub z}/PAN fibers were synthesized by in-situ method. • Photodegradation behavior of BiOCl{sub x}/BiOBr{sub y}/BiOI{sub z}/PAN fibers was measured under solar light irradiation. • BiOCl{sub 0.3}/BiOBr{sub 0.3}/BiOI{sub 0.4}/PAN fibers exhibited the highest photocatalytic activity. • Photocatalytic mechanism was discussed in detail. - Abstract: In this work, BiOCl{sub x}/BiOBr{sub y}/BiOI{sub z} (x + y + z = 1) composite nanofibers were prepared through electrospinning and the sol-gel methods. Photocatalytic degradation of trichloroethylene (TCE) by BiOCl{sub x}/BiOBr{sub y}/BiOI{sub z}/PAN nanofibers was systematically investigated via gas chromatography (GC). Optimum photocatalytic activity was achieved with BiOCl{sub 0.3}/BiOBr{sub 0.3}/BiOI{sub 0.4} fibers under solar light irradiation. X-ray photoelectron spectroscopy (XPS) peaks due to C−O and C=O were observed at 286.0 and 288.3 eV, respectively, it indicated that the BiOCl{sub x}/BiOBr{sub y}/BiOI{sub z} mixture had been successfully doped on the polyacrylonitrile (PAN) fibers. Furthermore, X-ray diffraction (XRD) results also confirmed that we had synthesized the as-prepared composite nanofibers successfully. Photocatalytic activities of BiOCl{sub 0.3}/BiOBr{sub 0.3}/BiOI{sub 0.4} were up to 3 times higher than the pure BiOCl, BiOBr and BiOI samples, respectively.

Results of 6th regular inspection of No.1 unit in Oi Power Plant

International Nuclear Information System (INIS)

Anon.

1987-01-01

This report presents results of the 6th regular inspection of the No.1 unit in the Oi Power Plant. It was carried out during the period from July 11, 1986, to January 28, 1987. The inspection covered the main unit of the nuclear reactor, facilities for the nuclear reactor cooling system, facilities for the instrumentation control system, fuel facilities, radiation control facilities, disposal facilities, nuclear reactor containment facilities, and emergency power generation system. Checking of appearance, disassemblage, leak and functions-performance of these facilities was conducted. No abnormalities were found except that significant signs were detected in 725 steam generator heat transfer pipes and that leak was suspected in 2 fuel assemblies. The pipes were repaired and the fuel assemblies were replaced. All operations involved in the inspection were performed under conditions within the permissible dose as specified in the applicable laws. Major modification work carried out during the inspection period included the adoption of a burnable poison (B Type) and the charging of fuel for high burn-up demonstration test. The exposure dose of the company members and non-company members who performed the inspection work is also shown. (Nogami, K.)

Facile Fabrication of BiOI/BiOCl Immobilized Films With Improved Visible Light Photocatalytic Performance

Directory of Open Access Journals (Sweden)

Yingxian Zhong

2018-03-01

Full Text Available HIGHLIGHTSA facial method was used to fabricate BiOI/BiOCl film at room temperature.30% BiOI/BiOCl showed an excellent photocatalytic activity and stability.Improvement of photocatalytic activity was owed to expanded visible light absorption and high separation efficiency of charge.Photocatalysis has been considered to be one of the most promising ways to photodegrade organic pollutants. Herein, a series of BiOI/BiOCl films coating on FTO were fabricated through a simple method at room temperature. The photocatalytic efficiency of 30%BiOI/BiOCl could reach more than 99% aiming to degrading RhB and MB after 90 and 120 min, respectively. Compared with BiOCl, 30%BiOI/BiOCl showed 12 times higher efficiency when degrading RhB. In comparison with BiOI, 30%BiOI/BiOCl showed 5 and 6 times higher efficiency when degrading RhB and MB, respectively. These obvious enhancements were attributed to expanded visible light absorption and high separation performance of photoinduced charge. Moreover, the photocatalytic activity of 30%BiOI/BiOCl had no obvious decrease after five recycles, suggesting that it was a promising photocatalyst for the removal of MB and RhB pollutants. Finally, the possible growth process for the BiOI/BiOCl thin films and photocatalysis mechanism were investigated in details. This work would provide insight to the reasonable construction of BiOX heterojunction and the photocatalytic mechanism in degrading organic pollutants.

Fabricaion of improved novel p–n junction BiOI/Bi{sub 2}Sn{sub 2}O{sub 7} nanocomposite for visible light driven photocatalysis

Energy Technology Data Exchange (ETDEWEB)

Xu, Weicheng [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Fang, Jianzhang, E-mail: fangjzh@scnu.edu.cn [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Guangdong Technology Research Center for Ecological Management and Remediation of Urban Water System, Guangzhou 510006 (China); Zhu, Ximiao [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Fang, Zhanqiang [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Guangdong Technology Research Center for Ecological Management and Remediation of Urban Water System, Guangzhou 510006 (China); Cen, Chaoping [The Key Laboratory of Water and Air Pollution Control of Guangdong Province, South China Institute of Environmental Sciences, Guangzhou 510655 (China)

2015-12-15

Graphical abstract: - Highlights: • A p–n heterojunction photocatalyst BiOI/Bi{sub 2}Sn{sub 2}O{sub 7} was prepared by hydrothermal method. • 4% BiOI/Bi{sub 2}Sn{sub 2}O{sub 7} with maximal photocatalytic degradation efficiency (RhB) of 99.9%. • A specific degradation routes of RhB was illustrated. • The photocatalytic mechanism is discussed according to p–n junction principles. • • O{sub 2}{sup −} and h+ are the main reactive species for the degradation of RhB. - Abstract: A series of novel p−n junction photocatalysts BiOI/Bi{sub 2}Sn{sub 2}O{sub 7} (BiOI/BSO) were successfully fabricated via a facile hydrothermal method. The phase structures, morphologies and optical properties of the as-prepared samples were studied by XRD, TEM, HRTEM, BET, XPS, UV–vis DRS and photoluminescence (PL) spectroscopy. The results showed that BiOI/BSO heteronanostructures displayed much higher photocatalytic activity than pure BSO and BiOI for the degradation of rhodamine B (RhB). The best photocatalytic activity of BiOI/BSO with almost 99.9% RhB degradation situated at molar percentage ratio of 4% after 6 h irradiation. The enhanced photocatalytic performance of BiOI/BSO could be mainly attributed to the formation of the heterojunction between p-BiOI and n-BSO, which effectively restrains the recombination of photoinduced electron–hole pairs. Moreover, the study of radical scavengers affirmed that h{sup +} and • O{sub 2}{sup −} were the primary reactive species for the degradation of RhB.

Facile Fabrication of BiOI/BiOCl Immobilized Films with Improved Visible Light Photocatalytic Performance

Science.gov (United States)

Zhong, Yingxian; Liu, Yuehua; Wu, Shuang; Zhu, Yi; Chen, Hongbin; Yu, Xiang; Zhang, Yuanming

2018-03-01

Photocatalysis has been considered to be one of the most promising ways to photodegrade organic pollutants. Herein, a series of BiOI/BiOCl films coating on FTO were fabricated through a simple method at room temperature. The photocatalytic efficiency of 30%BiOI/BiOCl could reach more than 99% aiming to degrading RhB and MB after 90 and 120 min, respectively. Compared with BiOCl, 30%BiOI/BiOCl showed 12 times higher efficiency when degrading RhB. In comparison with BiOI, 30%BiOI/BiOCl showed 5 and 6 times higher efficiency when degrading RhB and MB, respectively. These obvious enhancements were attributed to expanded visible light absorption and high separation performance of photoinduced charge. Moreover, the photocatalytic activity of 30%BiOI/BiOCl had no obvious decrease after 5 recycles, suggesting that it was a promising photocatalyst for the removal of MB and RhB pollutants. Finally, the possible growth process for the BiOI/BiOCl thin films and photocatalysis mechanism were investigated in details. This work would provide insight to the reasonable construction of BiOX heterojunction and the photocatalytic mechanism in degrading organic pollutants.

Tissue-specific mosaicism for a lethal osteogenesis imperfecta COL1A1 mutation causes mild OI/EDS overlap syndrome.

Science.gov (United States)

Symoens, Sofie; Steyaert, Wouter; Demuynck, Lynn; De Paepe, Anne; Diderich, Karin E M; Malfait, Fransiska; Coucke, Paul J

2017-04-01

Type I collagen is the predominant protein of connective tissues such as skin and bone. Mutations in the type I collagen genes (COL1A1 and COL1A2) mainly cause osteogenesis imperfecta (OI). We describe a patient with clinical signs of Ehlers-Danlos syndrome (EDS), including fragile skin, easy bruising, recurrent luxations, and fractures resembling mild OI. Biochemical collagen analysis of the patients' dermal fibroblasts showed faint overmodification of the type I collagen bands, a finding specific for structural defects in type I collagen. Bidirectional Sanger sequencing detected an in-frame deletion in exon 44 of COL1A1 (c.3150_3158del), resulting in the deletion of three amino acids (p.Ala1053_Gly1055del) in the collagen triple helix. This COL1A1 mutation was hitherto identified in four probands with lethal OI, and never in EDS patients. As the peaks on the electropherogram corresponding to the mutant allele were decreased in intensity, we performed next generation sequencing of COL1A1 to study mosaicism in skin and blood. While approximately 9% of the reads originating from fibroblast gDNA harbored the COL1A1 deletion, the deletion was not detected in gDNA from blood. Most likely, the mild clinical symptoms observed in our patient can be explained by the mosaic state of the mutation. © 2017 Wiley Periodicals, Inc.

NOAA Optimum Interpolation (OI) SST V2

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The optimum interpolation (OI) sea surface temperature (SST) analysis is produced weekly on a one-degree grid. The analysis uses in situ and satellite SST's plus...

Mode analysis and structure parameter optimization of a novel SiGe-OI rib optical waveguide

Energy Technology Data Exchange (ETDEWEB)

Feng Song; Gao Yong; Yang Yuan [Department of Electronic Engineering, Xi' an University of Technology, Xi' an 710048 (China); Feng Yuchun, E-mail: vonfs@yahoo.com.c [Key Laboratories of Optoelectronic Devices and Systems, Shenzhen University, Shenzhen 518060 (China)

2009-08-15

The mode of a novel SiGe-OI optical waveguide is analyzed, and its single-mode conditions are derived. The Ge content and structure parameters of SiGe-OI optical waveguides are respectively optimized. Under an operation wavelength of 1300 nm, the structures of SiGe-OI rib optical waveguides are built and analyzed with Optiwave software, and the optical field and transmission losses of the SiGe-OI rib optical waveguides are analyzed. The optimization results show that when the structure parameters H, h, W are respectively 500 nm, 250 nm, 500 nm and the Ge content is 5%, the total power loss of SiGe-OI rib waveguides is 0.3683 dB/cm considering the loss of radiation outside the waveguides and materials, which is less than the traditional value of 0.5 dB/cm. The analytical technique for SiGe-OI optical waveguides and structure parameters computed by this paper are proved to be accurate and computationally efficient compared with the beam propagation method (BPM) and the experimental results. (semiconductor devices)

Visible-Light-Driven BiOI-Based Janus Micromotor in Pure Water.

Science.gov (United States)

Dong, Renfeng; Hu, Yan; Wu, Yefei; Gao, Wei; Ren, Biye; Wang, Qinglong; Cai, Yuepeng

2017-02-08

Light-driven synthetic micro-/nanomotors have attracted considerable attention due to their potential applications and unique performances such as remote motion control and adjustable velocity. Utilizing harmless and renewable visible light to supply energy for micro-/nanomotors in water represents a great challenge. In view of the outstanding photocatalytic performance of bismuth oxyiodide (BiOI), visible-light-driven BiOI-based Janus micromotors have been developed, which can be activated by a broad spectrum of light, including blue and green light. Such BiOI-based Janus micromotors can be propelled by photocatalytic reactions in pure water under environmentally friendly visible light without the addition of any other chemical fuels. The remote control of photocatalytic propulsion by modulating the power of visible light is characterized by velocity and mean-square displacement analysis of optical video recordings. In addition, the self-electrophoresis mechanism has been confirmed for such visible-light-driven BiOI-based Janus micromotors by demonstrating the effects of various coated layers (e.g., Al 2 O 3 , Pt, and Au) on the velocity of motors. The successful demonstration of visible-light-driven Janus micromotors holds a great promise for future biomedical and environmental applications.

Transition from Pediatric to Adult OI Care

Science.gov (United States)

Moving from Pediatric to Adult Care Introduction Teen and young adult years are a critical time for major life changes. An ... for youth who have OI is moving from pediatric care into the adult care system. Children’s hospitals ...

Fuel recycling and 4. generation reactors

International Nuclear Information System (INIS)

Devezeaux de Lavergne, J.G.; Gauche, F.; Mathonniere, G.

2012-01-01

The 4. generation reactors meet the demand for sustainability of nuclear power through the saving of the natural resources, the minimization of the volume of wastes, a high safety standard and a high reliability. In the framework of the GIF (Generation 4. International Forum) France has decided to study the sodium-cooled fast reactor. Fast reactors have the capacity to recycle plutonium efficiently and to burn actinides. The long history of reprocessing-recycling of spent fuels in France is an asset. A prototype reactor named ASTRID could be entered into operation in 2020. This article presents the research program on the sodium-cooled fast reactor, gives the status of the ASTRID project and present the scenario of the progressive implementation of 4. generation reactors in the French reactor fleet. (A.C.)

François Louis (1928-2010)

CERN Multimedia

CERN Bulletin

2010-01-01

François Louis, who was a CERN mathematician from 1957 to 1988, passed away on 23 March 2010.   Everyone who was at CERN in the early years will remember the important role he played before the advent of computers, as well as his skills as a teacher when it became imperative for us all to make use of them. His lessons, at all levels, were remarkable for their precision and clarity. François loved to share his gifts of intelligence and culture with his friends and colleagues. Never turning down a request for help, he imparted his wisdom to the many who sought it with unfailing modesty and wit. We recall with affection his wonderful conversation, where mathematics merged with literature, cinema and his other great passion, music. Indeed he will also be remembered as a gifted pianist. His critical spirit and impeccable manners, born of a bygone era, left an indelible mark on us all. We are deeply saddened by his death. His friends      

Effective charge separation in BiOI/Cu2O composites with enhanced photocatalytic activity

Science.gov (United States)

Xia, Yongmei; He, Zuming; Yang, Wei; Tang, Bin; Lu, Yalin; Hu, Kejun; Su, Jiangbin; Li, Xiaoping

2018-02-01

Novel BiOI/Cu2O composites were designed and synthesized for the first time by coupling reduction method at low temperature. The samples were characterized by XRD, XPS, SEM, EDS, HRTEM, UV-vis (DRS), FTIR and photo-electro-chemical (PEC) analysis. Results showed that the BiOI/Cu2O composites consisted of three-dimensional (3D), hierarchical cauliflower-like structure composed of BiOI nanosheet and Cu2O cubic submicrometer structure, the composite absorption band broadened, and the absorption intensity in the visible region strengthened. And the composites exhibited an excellent photocatalytic performance, which might be attributed to the improvement of the composite absorption and effective charge separation in BiOI/Cu2O composites. In addition, the possible photocatalytic mechanism was proposed.

Photodegradation of Acid red 18 dye by BiOI/ZnO nanocomposite: A dataset

Directory of Open Access Journals (Sweden)

Sahand Jorfi

2018-02-01

Full Text Available Dyes are one of the most important existing pollutants in textile industrial wastewater. These compounds are often toxic, carcinogenic, and mutagenic to living organisms, chemically and photochemically stable, and non-biodegradable. Acid red 18 is one of the azo dyes that are currently used in the textile industries. Photocatalytic degradation offers a great potential as an advanced oxidation process, in this study photocatalytic degradation of Acid red 18 by using BiOI/ZnO nanocomposite was evaluated under visible light irradiation. The influence of most essential parameters such as pH and BiOI/ZnO dosage were studied for optimum conditions. The dye removal efficiency was 85.1% at optimum experimental conditions of pH of 7, and BiOI/ZnO dosage of 1.5 g/L. The data had a good agreement with pseudo first-order kinetic model. Thus, the BiOI/ZnO/UV is an efficient process for dye degradation. Keywords: Photodegradation, Nanocomposite, BiOI/ZnO, Degradation, Dye, Acid red 18

Diatomite-immobilized BiOI hybrid photocatalyst: Facile deposition synthesis and enhanced photocatalytic activity

International Nuclear Information System (INIS)

Li, Baoying; Huang, Hongwei; Guo, Yuxi; Zhang, Yihe

2015-01-01

Graphical abstract: - Highlights: • A novel diatomite-immobilized BiOI hybrid photocatalyst has been prepared by a facile one-step deposition process for the first time. • The diatomite-immobilized BiOI hybrid photocatalyst exhibits much better photocatalytic performance. • This enhancement should be attributed to that diatomite can play as an excellent carrier platform to increase the reactive sites and promote the separation of photogenerated electron–hole pairs. • This work shed new light on facile fabrication of novel composite photocatalyst based on natural mineral. - Abstract: A novel diatomite-immobilized BiOI hybrid photocatalyst has been prepared by a facile one-step deposition process for the first time. The structure, morphology and optical property of the products were characterized by X-ray powder diffraction (XRD), scanning electron microscopy (SEM) and UV–vis diffuse reflectance spectroscopy (DRS). The photocatalytic performance of the as-prepared BiOI/diatomite photocatalysts was studied by photodegradation of Rhodamine B (RhB) and methylene blue (MB) and monitoring photocurrent generation under visible light (λ > 420 nm). The results revealed that BiOI/diatomite composites exhibit enhanced photocatalytic activity compared to the pristine BiOI sample. This enhancement should be attributed to that diatomite can play as an excellent carrier platform to increase the reactive sites and promote the separation of photogenerated electron–hole pairs. In addition, the corresponding photocatalytic mechanism was proposed based on the active species trapping experiments. This work shed new light on facile fabrication of novel composite photocatalyst based on natural mineral.

Diatomite-immobilized BiOI hybrid photocatalyst: Facile deposition synthesis and enhanced photocatalytic activity

Energy Technology Data Exchange (ETDEWEB)

Li, Baoying; Huang, Hongwei, E-mail: hhw@cugb.edu.cn; Guo, Yuxi; Zhang, Yihe, E-mail: zyh@cugb.edu.cn

2015-10-30

Graphical abstract: - Highlights: • A novel diatomite-immobilized BiOI hybrid photocatalyst has been prepared by a facile one-step deposition process for the first time. • The diatomite-immobilized BiOI hybrid photocatalyst exhibits much better photocatalytic performance. • This enhancement should be attributed to that diatomite can play as an excellent carrier platform to increase the reactive sites and promote the separation of photogenerated electron–hole pairs. • This work shed new light on facile fabrication of novel composite photocatalyst based on natural mineral. - Abstract: A novel diatomite-immobilized BiOI hybrid photocatalyst has been prepared by a facile one-step deposition process for the first time. The structure, morphology and optical property of the products were characterized by X-ray powder diffraction (XRD), scanning electron microscopy (SEM) and UV–vis diffuse reflectance spectroscopy (DRS). The photocatalytic performance of the as-prepared BiOI/diatomite photocatalysts was studied by photodegradation of Rhodamine B (RhB) and methylene blue (MB) and monitoring photocurrent generation under visible light (λ > 420 nm). The results revealed that BiOI/diatomite composites exhibit enhanced photocatalytic activity compared to the pristine BiOI sample. This enhancement should be attributed to that diatomite can play as an excellent carrier platform to increase the reactive sites and promote the separation of photogenerated electron–hole pairs. In addition, the corresponding photocatalytic mechanism was proposed based on the active species trapping experiments. This work shed new light on facile fabrication of novel composite photocatalyst based on natural mineral.

High photocatalytic performance of BiOI/Bi{sub 2}WO{sub 6} toward toluene and Reactive Brilliant Red

Energy Technology Data Exchange (ETDEWEB)

Li Huiquan [School of Chemistry and Chemical Engineering, Fuyang Normal College, Fuyang 236041 (China); Key Laboratory of Mesoscopic Chemistry of MOE, Jiangsu Provincial Key Laboratory of Nanotechnology, School of Chemistry and Chemical Engineering, Nanjing University, Nanjing 210093 (China); Cui Yumin, E-mail: cuiyumin0908@163.com [School of Chemistry and Chemical Engineering, Fuyang Normal College, Fuyang 236041 (China); Hong Wenshan [School of Chemistry and Chemical Engineering, Fuyang Normal College, Fuyang 236041 (China)

2013-01-01

Graphical abstract: When BiOI/Bi{sub 2}WO{sub 6} catalyst was exposed to UV or visible light, the electrons in the valence band of Bi{sub 2}WO{sub 6} would be excited into the conduction band and then injected into the more positive conduction band of BiOI. Therefore, the photoelectrons were generated from Bi{sub 2}WO{sub 6} and transferred across the interface between BiOI and Bi{sub 2}WO{sub 6} to the surface of BiOI, leaving the photogenerated holes in the valence band of Bi{sub 2}WO{sub 6}. In this way, the photoinduced electron-hole pairs were effectively separated. Highlights: Black-Right-Pointing-Pointer BiOI sensitized Bi{sub 2}WO{sub 6} catalysts were successfully prepared by a facile method. Black-Right-Pointing-Pointer The 13.2% BiOI/Bi{sub 2}WO{sub 6} catalyst exhibits higher photoactivities than P25. Black-Right-Pointing-Pointer A possible transfer process of photogenerated carriers was proposed. - Abstract: BiOI sensitized nano-Bi{sub 2}WO{sub 6} photocatalysts with different BiOI contents were successfully synthesized by a facile deposition method at room temperature, and characterized by X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS) high-resolution transmission electron microscopy (HR-TEM), photoluminescence (PL) spectra, UV-vis diffuse reflection spectroscopy (UV-vis DRS) and Brunauer-Emmett-Teller (BET) surface area measurements. The photocatalytic activity of BiOI/Bi{sub 2}WO{sub 6} was evaluated by the photo-degradation of Reactive Brilliant Red (X-3B) in suspended solution and toluene in gas phase. It has been shown that the BiOI/Bi{sub 2}WO{sub 6} catalysts exhibit a coexistence of both tetragonal BiOI and orthorhombic Bi{sub 2}WO{sub 6} phases. With increasing BiOI content, the absorption intensity of BiOI/Bi{sub 2}WO{sub 6} catalysts increases in the 380-600 nm region and the absorption edge shifts significantly to longer wavelengths as compared to pure Bi{sub 2}WO{sub 6}. The 13.2% BiOI/Bi{sub 2}WO{sub 6} catalyst exhibits

Diatomite-immobilized BiOI hybrid photocatalyst: Facile deposition synthesis and enhanced photocatalytic activity

Science.gov (United States)

Li, Baoying; Huang, Hongwei; Guo, Yuxi; Zhang, Yihe

2015-10-01

A novel diatomite-immobilized BiOI hybrid photocatalyst has been prepared by a facile one-step deposition process for the first time. The structure, morphology and optical property of the products were characterized by X-ray powder diffraction (XRD), scanning electron microscopy (SEM) and UV-vis diffuse reflectance spectroscopy (DRS). The photocatalytic performance of the as-prepared BiOI/diatomite photocatalysts was studied by photodegradation of Rhodamine B (RhB) and methylene blue (MB) and monitoring photocurrent generation under visible light (λ > 420 nm). The results revealed that BiOI/diatomite composites exhibit enhanced photocatalytic activity compared to the pristine BiOI sample. This enhancement should be attributed to that diatomite can play as an excellent carrier platform to increase the reactive sites and promote the separation of photogenerated electron-hole pairs. In addition, the corresponding photocatalytic mechanism was proposed based on the active species trapping experiments. This work shed new light on facile fabrication of novel composite photocatalyst based on natural mineral.

Hommage à François Morel (1923-2007)

OpenAIRE

Corvol, Pierre

2010-01-01

Le 9 mai 2007, François Morel nous quittait à l’âge de 84 ans, après toute une vie consacrée à la recherche sur la physiologie rénale qu’il a réalisée en grande partie au Collège de France où il a été titulaire de la chaire de Physiologie cellulaire de 1967 à 1993. François Morel était né à Genève en 1923. Son père était titulaire de la chaire de Psychiatrie dans cette ville, ce qui l’a sans doute incité à entreprendre des études de médecine. En fait, il n’a jamais exercé la médecine car, trè...

Social and Emotional Issues of Living with OI

Science.gov (United States)

... Unbreakable Spirit® OI Golf Classic Awareness Week Fine Wines Strong Bones Bone China Tea Blue Jeans for ... children. Social skills learned as a child will benefit the teen and young adult. Joining teams, clubs, ...

Validity and reliability of the Turkish version of the Optimality Index-US (OI-US) to assess maternity care outcomes.

Science.gov (United States)

Yucel, Cigdem; Taskin, Lale; Low, Lisa Kane

2015-12-01

Although obstetrical interventions are used commonly in Turkey, there is no standardized evidence-based assessment tool to evaluate maternity care outcomes. The Optimality Index-US (OI-US) is an evidence-based tool that was developed for the purpose of measuring aggregate perinatal care processes and outcomes against an optimal or best possible standard. This index has been validated and used in Netherlands, USA and UK until now. The objective of this study was to adapt the OI-US to assess maternity care outcomes in Turkey. Translation and back translation were used to develop the Optimality Index-Turkey (OI-TR) version. To evaluate the content validity of the OI-TR, an expert panel group (n=10) reviewed the items and evidence-based quality of the OI-TR for application in Turkey. Following the content validity process, the OI-TR was used to assess 150 healthy and 150 high-risk pregnant women who gave birth at a high volume, urban maternity hospital in Turkey. The scores between the two groups were compared to assess the discriminant validity of the OI-TR. The percentage of agreement between two raters and the Kappa statistic were calculated to evaluate the reliability. Content validity was established for the OI-TR by an expert group. Discriminant validity was confirmed by comparing the OI scores of healthy pregnant women (mean OI score=77.65%) and those of high-risk pregnant women (mean OI score=78.60%). The percentage of agreement between the two raters was 96.19, and inter-rater agreement was provided for each item in the OI-TR. OI-TR is a valid and reliable tool that can be used to assess maternity care outcomes in Turkey. The results of this study indicate that although the risk statuses of the women differed, the type of care they received was essentially the same, as measured by the OI-TR. Care was not individualised based on risk and for a majority of items was inconsistent with evidence based practice, which is not optimal. Use of the OI-TR will help to

Study on gas-liquid loop reactors with annular bubbling

International Nuclear Information System (INIS)

Fei, L.M.; Wang, S.X.; Wu, X.Q.; Lu, D.W.

1987-01-01

Bubbling column with draft tube is one of nearly developed reactor. On the background of hydrocarbon oxidations and biochemical engineerings, it has been widely used in chemical industry due to the well characteristics of mass and heat transfer. In this paper, the characteristics of fluid flow, such as gas hold-up, backmixing and mass transfer referred to the liquid volume were measured in a gas-liquid loop reactor with annular bubbling. Different materials - water, alcohol and oi l- were used in the study in measuring the gas hold-up in the annular of the reactor

Heterojunctions of p-BiOI Nanosheets/n-TiO2 Nanofibers: Preparation and Enhanced Visible-Light Photocatalytic Activity

Directory of Open Access Journals (Sweden)

Kexin Wang

2016-01-01

Full Text Available p-BiOI nanosheets/n-TiO2 nanofibers (p-BiOI/n-TiO2 NFs have been facilely prepared via the electrospinning technique combining successive ionic layer adsorption and reaction (SILAR. Dense BiOI nanosheets with good crystalline and width about 500 nm were uniformly assembled on TiO2 nanofibers at room temperature. The amount of the heterojunctions and the specific surface area were well controlled by adjusting the SILAR cycles. Due to the synergistic effect of p-n heterojunctions and high specific surface area, the obtained p-BiOI/n-TiO2 NFs exhibited enhanced visible-light photocatalytic activity. Moreover, the p-BiOI/n-TiO2 NFs heterojunctions could be easily recycled without decreasing the photocatalytic activity owing to their one-dimensional nanofibrous structure. Based on the above, the heterojunctions of p-BiOI/n-TiO2 NFs may be promising visible-light-driven photocatalysts for converting solar energy to chemical energy in environment remediation.

Particules de vie conversation avec François Englert

CERN Document Server

Baré, François

2014-01-01

Ce fut l'événement de l'année 2013 : François Englert obtenait le prix Nobel de physique pour la découverte du boson. Le premier Belge à être honoré par ce prix depuis Ilya Prigogine en 1977. Françoise Baré, journaliste à la RTBF, et Guy Duplat, ingénieur civil physicien, ex-rédacteur en chef du Soir, auteur de Une vague belge (Racine) et journaliste à La Libre Belgique, ont rencontré plusieurs fois, longuement, François Englert, avant son prix et après celui-ci. Ils étaient là lors de l'annonce de la découverte du boson de BEH, lors de l'annonce du prix Nobel et, à Stockholm, lors de la remise du prix par le roi de Suède. À eux, François Englert a accepté de raconter sa chasse au boson, sa vie, ses idées. Au fil de la conversation se dessine alors le portrait attachant d'un homme brillant et libre, habité par un rêve, celui de comprendre, de décortiquer ce monde et de chercher la beauté de ses lois. Avec, en annexe, les grands points scientifiques qui éclairent le travail de F...

Facile synthesis of flower-like BiOI hierarchical spheres at room temperature with high visible-light photocatalytic activity

International Nuclear Information System (INIS)

Wang, Xiao-jing; Li, Fa-tang; Li, Dong-yan; Liu, Rui-hong; Liu, Shuang-jun

2015-01-01

Graphical abstract: - Highlights: • Flower-like BiOI hierarchical sphere is obtained in the presence of ethylene glycol. • A template free hydrolysis route is employed at room temperature. • Ethylene glycol plays an important role in assembling BiOI nanoflakes to form spheres. • The BiOI sphere shows high visible-light photocatalytic activity and good stability. - Abstract: Flower-like BiOI hierarchical spheres are prepared at room temperature via a template free route simply by dropping water into ethylene glycol (EG) solution containing reactants based on the hydrolysis and oriented assembly roles of water and EG, respectively. The BiOI samples are characterized by X-ray diffraction (XRD), nitrogen adsorption/desorption, emission scanning electron microscopy (SEM), UV–Vis diffuse reflectance spectra (UV–Vis DRS), X-ray photoelectron spectroscopy (XPS), and transmission electron microscopy (TEM). The photocatalytic reaction rate constant of the as-prepared BiOI hierarchical spheres is 15.8, 13.3, and 2.0 times that of BiOI nanoflakes obtained in the absence of EG in degradation of anionic dye (methyl orange), cationic dye (methylene blue), and colorless target pollutant (phenol), respectively, under the visible-light irradiation, which can be attributed to its unique flower-like structure for utilization of light, small crystal size, and large specific surface area

Decoration of BiOI quantum size nanoparticles with reduced graphene oxide in enhanced visible-light-driven photocatalytic studies

International Nuclear Information System (INIS)

Liu Zhang; Xu Weicheng; Fang Jianzhang; Xu Xiaoxin; Wu Shuxing; Zhu Ximiao; Chen Zehua

2012-01-01

Highlights: ► RGO/BiOI nanocomposites were synthesized by a reverse microemulsion method. ► Quantum sized BiOI nanoparticles can be obtained by this approach. ► Ascorbic acid was used as a reducing agent to reduce GO and seemed to be effective. ► RGO/BiOI presented outstanding visible-light-induced photocatalytic performance. ► Possible photocatalytic mechanism was proposed based on the experimental studies. - Abstract: Herein, a reverse microemulsion route was developed to synthesize bismuth oxyiodide (BiOI) nanocrystals and reduced graphene oxide (RGO) nanocomposites as a highly efficient photocatalyst, and both the formation of BiOI and the reduction of RGO were achieved in situ in microemulsions simultaneously at low temperature (60 °C). The uniform nanocrystal size and structure were indicated by XRD, TEM, and the reduction of GO by ascorbic acid was evidenced by FTIR, XPS, and Raman spectra techniques. The enhanced photoactivity of RGO/BiOI nanocomposites under visible light was attributed to improved light absorption and efficient charge separation and transportation.

Decoration of BiOI quantum size nanoparticles with reduced graphene oxide in enhanced visible-light-driven photocatalytic studies

Energy Technology Data Exchange (ETDEWEB)

Liu Zhang, E-mail: liuzhang0126@126.com [School of Chemistry and Environment, South China Normal University, Guangzhou 510006 (China); Xu Weicheng [School of Chemistry and Environment, South China Normal University, Guangzhou 510006 (China); Fang Jianzhang, E-mail: fangjzh@scnu.edu.cn [School of Chemistry and Environment, South China Normal University, Guangzhou 510006 (China); Xu Xiaoxin; Wu Shuxing; Zhu Ximiao; Chen Zehua [School of Chemistry and Environment, South China Normal University, Guangzhou 510006 (China)

2012-10-15

Highlights: Black-Right-Pointing-Pointer RGO/BiOI nanocomposites were synthesized by a reverse microemulsion method. Black-Right-Pointing-Pointer Quantum sized BiOI nanoparticles can be obtained by this approach. Black-Right-Pointing-Pointer Ascorbic acid was used as a reducing agent to reduce GO and seemed to be effective. Black-Right-Pointing-Pointer RGO/BiOI presented outstanding visible-light-induced photocatalytic performance. Black-Right-Pointing-Pointer Possible photocatalytic mechanism was proposed based on the experimental studies. - Abstract: Herein, a reverse microemulsion route was developed to synthesize bismuth oxyiodide (BiOI) nanocrystals and reduced graphene oxide (RGO) nanocomposites as a highly efficient photocatalyst, and both the formation of BiOI and the reduction of RGO were achieved in situ in microemulsions simultaneously at low temperature (60 Degree-Sign C). The uniform nanocrystal size and structure were indicated by XRD, TEM, and the reduction of GO by ascorbic acid was evidenced by FTIR, XPS, and Raman spectra techniques. The enhanced photoactivity of RGO/BiOI nanocomposites under visible light was attributed to improved light absorption and efficient charge separation and transportation.

Sodium citrate-assisted anion exchange strategy for construction of Bi2O2CO3/BiOI photocatalysts

International Nuclear Information System (INIS)

Song, Peng-Yuan; Xu, Ming; Zhang, Wei-De

2015-01-01

Highlights: • Heterostructured Bi 2 O 2 CO 3 /BiOI microspheres were prepared via anion exchange. • Sodium citrate-assisted anion exchange for construction of composite photocatalysts. • Bi 2 O 2 CO 3 /BiOI composites show high visible light photocatalytic activity. - Abstract: Bi 2 O 2 CO 3 /BiOI heterojuncted photocatalysts were constructed through a facile partial anion exchange strategy starting from BiOI microspheres and urea with the assistance of sodium citrate. The content of Bi 2 O 2 CO 3 in the catalysts was regulated by modulating the amount of urea as a precursor, which was decomposed to generate CO 3 2− in the hydrothermal process. Citrate anion plays a key role in controlling the morphology and composition of the products. The Bi 2 O 2 CO 3 /BiOI catalysts display much higher photocatalytic activity than pure BiOI and Bi 2 O 2 CO 3 towards the degradation of rhodamine B (RhB) and bisphenol A (BPA). The enhancement of photocatalytic activity of the heterojuncted catalysts is attributed to the formation of p–n junction between p-BiOI and n-Bi 2 O 2 CO 3 , which is favorable for retarding the recombination of photoinduced electron-hole pairs. Moreover, the holes are demonstrated to be the main active species for the degradation of RhB and BPA

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

Monte Carlo modeling of Io’s [OI] 6300 Å and [SII] 6716 Å auroral emission in eclipse

Science.gov (United States)

Moore, C.; Miki, K.; Goldstein, D. B.; Stapelfeldt, K.; Varghese, P. L.; Trafton, L. M.; Evans, R. W.

2010-06-01

We present a Monte Carlo (MC) model of [OI] 6300 Å and [SII] 6716 Å emission from Io entering eclipse. The simulation accounts for the 3-D distribution of SO 2, O, SO, S, and O 2 in Io's atmosphere, several volcanic plumes, and the magnetic field around Io. Thermal electrons from the jovian plasma torus are input along the simulation domain boundaries and move along the magnetic field lines distorted by Io, occasionally participating in collisions with neutrals. We find that the atmospheric asymmetry resulting from varying degrees of atmospheric collapse across Io (due to eclipse ingress) and the presence of volcanoes contributes significantly to the unique morphology of the [OI] 6300 Å emission. The [OI] radiation lifetime of ˜134 s limits the emission to regions that have a sufficiently low neutral density so that intermolecular collisions are rare. We find that at low altitudes (typically Pele, Prometheus, etc.) the number density is large enough (>4 × 10 9 cm -3) to collisionally quench nearly all (>95%) of the excited oxygen for reasonable quenching efficiencies. Upstream (relative to the plasma flow), Io's perturbation of the jovian magnetic field mirrors electrons with high pitch angles, while downstream collisions can trap the electrons. This magnetic field perturbation is one of the main physical mechanisms that results in the upstream/downstream brightness asymmetry in [OI] emission seen in the observation by Trauger et al. (Trauger, J.T., Stapelfeldt, K.R., Ballester, G.E., Clarke, J.I., 1997. HST observations of [OI] emissions from Io in eclipse. AAS-DPS Abstract (1997DPS29.1802T)). There are two other main causes for the observed brightness asymmetry. First, the observation's viewing geometry of the wake spot crosses the dayside atmosphere and therefore the wake's observational field of view includes higher oxygen column density than the upstream side. Second, the phased entry into eclipse results in less atmospheric collapse and thus higher

DFT study on the interfacial properties of vertical and in-plane BiOI/BiOIO3 hetero-structures.

Science.gov (United States)

Dai, Wen-Wu; Zhao, Zong-Yan

2017-04-12

Composite photocatalysts with hetero-structures usually favor the effective separation of photo-generated carriers. In this study, BiOIO 3 was chosen to form a hetero-structure with BiOI, due to its internal polar field and good lattice matching with BiOI. The interfacial properties and band offsets were focused on and analyzed in detail by DFT calculations. The results show that the charge depletion and accumulation mainly occur in the region near the interface. This effect leads to an interfacial electric field and thus, the photo-generated electron-hole pairs can be easily separated and transferred along opposite directions at the interface, which is significant for the enhancement of the photocatalytic activity. Moreover, according to the analysis of band offsets, the vertical BiOI/BiOIO 3 belongs to the type-II hetero-structure, while the in-plane BiOI/BiOIO 3 belongs to the type-I hetero-structure. The former type of hetero-structure has more favorable effects to enhance the photocatalytic activity of BiOI than that of the latter type of hetero-structure. In the case of the vertical BiOI/BiOIO 3 hetero-structure, photo-generated electrons can move from the conduction band of BiOI to that of BiOIO 3 , while holes can move from the valence band of BiOIO 3 to that of BiOI under solar radiation. In addition, the introduced internal electric field functions as a selector that can promote the separation of photo-generated carriers, resulting in the higher photocatalytic quantum efficiency. These findings illustrate the underlying mechanism for the reported experiments, and can be used as a basis for the design of novel highly efficient composite photocatalysts with hetero-structures.

Ti(Oi-Pr)4-promoted photoenolization Diels-Alder reaction to construct polycyclic rings and its synthetic applications.

Science.gov (United States)

Yang, Baochao; Lin, Kuaikuai; Shi, Yingbo; Gao, Shuanhu

2017-09-20

Stereoselective construction of polycyclic rings with all-carbon quaternary centers, and vicinal all-carbon quaternary stereocenters, remains a significant challenge in organic synthesis. These structures can be found in a wide range of polycyclic natural products and drug molecules. Here we report a Ti(Oi-Pr) 4 -promoted photoenolization/Diels-Alder (PEDA) reaction to construct hydroanthracenol and related polycyclic rings bearing all-carbon quaternary centers. This photolysis proceeds under mild conditions and generates a variety of photo-cycloaddition products in good reaction efficiency and stereoselectivity (48 examples), and has been successfully used in the construction of core skeleton of oncocalyxones, tetracycline and pleurotin. It also provides a reliable method for the late-stage modification of natural products bearing enone groups, such as steroids. The total synthesis of oncocalyxone B was successfully achieved using this PEDA approach.Anthracenols with multiple chiral centres are common motifs in natural products. Here, the authors show a highly stereoselective photoenolization/Diels-Alder methodology involving a key Lewis acid reagent enabling the efficient construction of a family of anthracenol derivatives with quaternary centers.

Facile synthesis of AgI/BiOI-Bi{sub 2}O{sub 3} multi-heterojunctions with high visible light activity for Cr(VI) reduction

Energy Technology Data Exchange (ETDEWEB)

Wang, Qi [School of Environmental Science and Engineering, Zhejiang Gongshang University, Hangzhou 310018 (China); The Brook Byer Institute for Sustainable Systems and School of Civil and Environmental Engineering, Georgia Institute of Technology, Atlanta 30332 (United States); Shi, Xiaodong; Liu, Enqin [School of Environmental Science and Engineering, Zhejiang Gongshang University, Hangzhou 310018 (China); Crittenden, John C. [The Brook Byer Institute for Sustainable Systems and School of Civil and Environmental Engineering, Georgia Institute of Technology, Atlanta 30332 (United States); Ma, Xiangjuan; Zhang, Yi [School of Environmental Science and Engineering, Zhejiang Gongshang University, Hangzhou 310018 (China); Cong, Yanqing, E-mail: yqcong@hotmail.com [School of Environmental Science and Engineering, Zhejiang Gongshang University, Hangzhou 310018 (China)

2016-11-05

Graphical abstract: Highly visible-light-active AgI/BiOI-Bi{sub 2}O{sub 3} with multi-heterojunctions was developed. - Highlights: • Visible-light-active AgI/BiOI-Bi{sub 2}O{sub 3} with multi-heterojunctions was prepared. • Highly enhanced photocatalytic reduction of Cr(VI) was observed. • k{sub Cr(VI)} on AgI/BiOI-Bi{sub 2}O{sub 3} increased by ca.16 times relative to Bi{sub 2}O{sub 3}. • Decreased E{sub g}, shifted E{sub fb} and reduced charge transfer resistance were observed. • Simultaneous reduction of Cr(VI) and degradation of organics were achieved. - Abstract: AgI sensitized BiOI-Bi{sub 2}O{sub 3} composite (AgI/BiOI-Bi{sub 2}O{sub 3}) with multi-heterojunctions was prepared using simple etching-deposition process. Different characterization techniques were performed to investigate the structural, optical and electrical properties of the as-prepared photocatalysts. It was found that the ternary AgI/BiOI-Bi{sub 2}O{sub 3} composite exhibited: (1) improved photocurrent response, (2) smaller band gap, (3) greatly reduced charge transfer resistance and (4) negative shift of flat band potential, which finally led to easier generation and more efficient separation of photo-generated electron-hole pairs at the hetero-interfaces. Thus, for the reduction of Cr(VI), AgI/BiOI-Bi{sub 2}O{sub 3} exhibited excellent photocatalytic activity under visible light irradiation at near neutral pH. AgI/BiOI-Bi{sub 2}O{sub 3} was optimized when the initial molar ratio of KI to Bi{sub 2}O{sub 3} and AgNO{sub 3} to Bi{sub 2}O{sub 3} was 1:1 and 10%, respectively. The estimated k{sub Cr(VI)} on optimized AgI/BiOI-Bi{sub 2}O{sub 3} was about 16 times that on pure Bi{sub 2}O{sub 3}. Good stability was also observed in cyclic runs, indicating that the current multi-heterostructured photocatalyst is highly desirable for the remediation of Cr(VI)-containing wastewater.

Simultaneous imaging of aurora on small scale in OI (777.4 nm and N21P to estimate energy and flux of precipitation

Directory of Open Access Journals (Sweden)

N. Ivchenko

2009-07-01

Full Text Available Simultaneous images of the aurora in three emissions, N21P (673.0 nm, OII (732.0 nm and OI (777.4 nm, have been analysed; the ratio of atomic oxygen to molecular nitrogen has been used to provide estimates of the changes in energy and flux of precipitation within scale sizes of 100 m, and with temporal resolution of 32 frames per second. The choice of filters for the imagers is discussed, with particular emphasis on the choice of the atomic oxygen line at 777.4 nm as one of the three emissions measured. The optical measurements have been combined with radar measurements and compared with the results of an auroral model, hence showing that the ratio of emission rates OI/N2 can be used to estimate the energy within the smallest auroral structures. In the event chosen, measurements were made from mainland Norway, near Tromso, (69.6 N, 19.2 E. The peak energies of precipitation were between 1–15 keV. In a narrow curling arc, it was found that the arc filaments resulted from energies in excess of 10 keV and fluxes of approximately 7 mW/m2. These filaments of the order of 100 m in width were embedded in a region of lower energies (about 5–10 keV and fluxes of about 3 mW/m2. The modelling results show that the method promises to be most powerful for detecting low energy precipitation, more prevalent at the higher latitudes of Svalbard where the multispectral imager, known as ASK, is now installed.

Simultaneous imaging of aurora on small scale in OI (777.4 nm and N21P to estimate energy and flux of precipitation

Directory of Open Access Journals (Sweden)

B. S. Lanchester

2009-07-01

Full Text Available Simultaneous images of the aurora in three emissions, N21P (673.0 nm, OII (732.0 nm and OI (777.4 nm, have been analysed; the ratio of atomic oxygen to molecular nitrogen has been used to provide estimates of the changes in energy and flux of precipitation within scale sizes of 100 m, and with temporal resolution of 32 frames per second. The choice of filters for the imagers is discussed, with particular emphasis on the choice of the atomic oxygen line at 777.4 nm as one of the three emissions measured. The optical measurements have been combined with radar measurements and compared with the results of an auroral model, hence showing that the ratio of emission rates OI/N2 can be used to estimate the energy within the smallest auroral structures. In the event chosen, measurements were made from mainland Norway, near Troms\\o, (69.6 N, 19.2 E. The peak energies of precipitation were between 1–15 keV. In a narrow curling arc, it was found that the arc filaments resulted from energies in excess of 10 keV and fluxes of approximately 7 mW/m2. These filaments of the order of 100 m in width were embedded in a region of lower energies (about 5–10 keV and fluxes of about 3 mW/m2. The modelling results show that the method promises to be most powerful for detecting low energy precipitation, more prevalent at the higher latitudes of Svalbard where the multispectral imager, known as ASK, is now installed.

Exercise and Activity: Key Elements in the Management of OI

Science.gov (United States)

... with peers. Children and adults with OI will benefit from a regular program of physical activity to promote optimal function through muscle strengthening, aerobic exercise, and recreational pursuits. Specifics of the exercise program vary depending ...

One-pot solvothermal synthesis of three-dimensional (3D) BiOI/BiOCl composites with enhanced visible-light photocatalytic activities for the degradation of bisphenol-A

International Nuclear Information System (INIS)

Xiao, Xin; Hao, Rong; Liang, Min; Zuo, Xiaoxi; Nan, Junmin; Li, Laisheng; Zhang, Weide

2012-01-01

Highlights: ► Synthesis of 3D BiOI/BiOCl microspheres by a one-pot template-free solvothermal method. ► Photocatalyst is BiOI/BiOCl composites. ► BiOI/BiOCl composites have enhanced visible-light photocatalytic ability to bisphenol-A. ► A simple and direct photodegradation pathway of bisphenol-A is proposed. - Abstract: Three-dimensional (3D) BiOI/BiOCl composite microspheres with enhanced visible-light photodegradation activity of bisphenol-A (BPA) are synthesized by a simple, one-pot, template-free, solvothermal method using BiI 3 and BiCl 3 as precursors. These 3D hierarchical microspheres with heterojunction structures are composed of 2D nanosheets and have composition-dependent absorption properties in the ultraviolet and visible light regions. The photocatalytic oxidation of BPA over BiOI/BiOCl composites followed pseudo first-order kinetics according to the Langmuir–Hinshelwood model. The highest photodegradation efficiency of BPA, i.e., nearly 100%, was observed with the BiOI/BiOCl composite (containing 90% BiOI) using a catalyst dosage of 1 g L −1 in the BPA solution (C 0 = 20 mg L −1 , pH = 7.0) under visible light irradiation for 60 min. Under these conditions, the reaction rate constant was more than 4 and 20 times greater than that of pure BiOI and the commercially available Degussa P25, respectively. The superior photocatalytic activity of this composite catalyst is attributed to the suitable band gap energies and the low recombination rate of the photogenerated electron–hole pairs due to the presence of BiOI/BiOCl heterostructures. Only one intermediate at m/z 151 was observed in the photodegradation process of BPA by liquid chromatography combined with mass spectrometry (LC–MS) analysis, and a simple and hole-predominated photodegradation pathway of BPA was subsequently proposed. Furthermore, this photocatalyst exhibited a high mineralization ratio, high stability and easy separation for recycling use, suggesting that it is a

François Delsarte and Modern Dance: an encounter in physical expression

Directory of Open Access Journals (Sweden)

Elisa Teixeira de Souza

2012-11-01

Full Text Available This study addresses François Delsarte’s system of expression, known as Applied Aesthetics. It presents data related to François Delsarte’s career, such as personal and professional life and his theoretical background. It discusses the laws of gestural expression formulated by Delsarte – Trinity Law, the Law of Correspondence and the Nine Laws of Motion – as well as their dissemination and utilization in modern dance; this discussion mentions some pioneers of modern dance, such as Isadora Duncan, Ruth Saint Denis, Ted Shawn, Vaslav Nijinsky, Rudolf Laban and Mary Wigman.

Two dimensional visible-light-active Pt-BiOI photoelectrocatalyst for efficient ethanol oxidation reaction in alkaline media

Science.gov (United States)

Zhai, Chunyang; Hu, Jiayue; Sun, Mingjuan; Zhu, Mingshan

2018-02-01

Two dimensional (2D) BiOI nanoplates were synthesized and used as support for the deposition of Pt nanoparticles. Owing to broad visible light absorption (up to 660 nm), the as-obtained Pt-BiOI electrode was used as effective photoelectrocatalyst in the application of catalytic ethanol oxidation in alkaline media under visible light irradiation. Compared to dark condition, the Pt-BiOI modified electrode displayed 3 times improved catalytic activity towards ethanol oxidation under visible light irradiation. The synergistic effect of electrocatalytic and photocatalytic, and the unique of 2D structures contribute to the improvement of catalytic activity. The mechanism of enhanced photoelectrocatalytic process is proposed. The present results suggest that 2D visible-light-activated BiOI can be served as promising support for the decoration of Pt and applied in the fields of photoelectrochemical and photo-assisted fuel cell applications

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

International Nuclear Information System (INIS)

1984-01-01

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

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

OI Fluorescent Line Contamination in Soft X-Ray Diffuse Background Obtained with Suzaku/XIS

OpenAIRE

Sekiya, Norio; Yamasaki, Noriko Y.; Mitsuda, Kazuhisa; Takei, Yoh

2014-01-01

The quantitative measurement of OVII line intensity is a powerful method for understanding the soft X-ray diffuse background. By systematically analyzing the OVII line intensity in 145 high-latitude Suzaku/XIS observations, the flux of OI fluorescent line in the XIS spectrum, contaminating the OVII line, is found to have an increasing trend with time especially after 2011. For these observations, the OVII line intensity would be overestimated unless taking into consideration the OI fluorescen...

Results of 7th regular inspection of No.1 plant in Oi Power Station, Kansai Electric Power Co., Inc

International Nuclear Information System (INIS)

1989-01-01

The 7th regular inspection of No.1 plant in Oi Power Station was carried out from December 25, 1987 to July 15, 1988. The parallel operation was resumed on June 23, 1988, 182 days after the start of the inspection. The facilities to be inspected were the reactor proper, reactor cooling system, measurement and control system, fuel facilities, radiation control facilities, waste facilities, reactor containment installation and emergency power generation system. On these facilities to be inspected, the appearance, disassembling, leak, function, performance and other inspections were carried out, and as the results, a part of the fitting of a water chamber partition cover for a steam generator broke off, significant signals were observed in 936 heating tubes of steam generators, 72 bolts for fixing the blade of a primary coolant pump were damaged, and leak was found in two fuel assemblies. The works related to this regular inspection were accomplished within the range of allowable radiation dose based on the relevant laws. The main reconstruction works carried out during the period of this regular inspection were the use of the fuel containing gadolinia, the removal of a thermometer bypass piping and the repair of defective steam generator tubes. (K.I.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-01

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

One-pot solvothermal synthesis of three-dimensional (3D) BiOI/BiOCl composites with enhanced visible-light photocatalytic activities for the degradation of bisphenol-A

Energy Technology Data Exchange (ETDEWEB)

Xiao, Xin [School of Chemistry and Environment, South China Normal University, Key Lab of Theoretical Chemistry of Environment, Guangzhou 510006 (China); Nano Science Research Center, School of Chemistry and Chemical Engineering, South China University of Technology, Guangzhou 510640 (China); Hao, Rong; Liang, Min; Zuo, Xiaoxi [School of Chemistry and Environment, South China Normal University, Key Lab of Theoretical Chemistry of Environment, Guangzhou 510006 (China); Nan, Junmin, E-mail: jmnan@scnu.edu.cn [School of Chemistry and Environment, South China Normal University, Key Lab of Theoretical Chemistry of Environment, Guangzhou 510006 (China); Li, Laisheng [School of Chemistry and Environment, South China Normal University, Key Lab of Theoretical Chemistry of Environment, Guangzhou 510006 (China); Zhang, Weide [Nano Science Research Center, School of Chemistry and Chemical Engineering, South China University of Technology, Guangzhou 510640 (China)

2012-09-30

Highlights: Black-Right-Pointing-Pointer Synthesis of 3D BiOI/BiOCl microspheres by a one-pot template-free solvothermal method. Black-Right-Pointing-Pointer Photocatalyst is BiOI/BiOCl composites. Black-Right-Pointing-Pointer BiOI/BiOCl composites have enhanced visible-light photocatalytic ability to bisphenol-A. Black-Right-Pointing-Pointer A simple and direct photodegradation pathway of bisphenol-A is proposed. - Abstract: Three-dimensional (3D) BiOI/BiOCl composite microspheres with enhanced visible-light photodegradation activity of bisphenol-A (BPA) are synthesized by a simple, one-pot, template-free, solvothermal method using BiI{sub 3} and BiCl{sub 3} as precursors. These 3D hierarchical microspheres with heterojunction structures are composed of 2D nanosheets and have composition-dependent absorption properties in the ultraviolet and visible light regions. The photocatalytic oxidation of BPA over BiOI/BiOCl composites followed pseudo first-order kinetics according to the Langmuir-Hinshelwood model. The highest photodegradation efficiency of BPA, i.e., nearly 100%, was observed with the BiOI/BiOCl composite (containing 90% BiOI) using a catalyst dosage of 1 g L{sup -1} in the BPA solution (C{sub 0} = 20 mg L{sup -1}, pH = 7.0) under visible light irradiation for 60 min. Under these conditions, the reaction rate constant was more than 4 and 20 times greater than that of pure BiOI and the commercially available Degussa P25, respectively. The superior photocatalytic activity of this composite catalyst is attributed to the suitable band gap energies and the low recombination rate of the photogenerated electron-hole pairs due to the presence of BiOI/BiOCl heterostructures. Only one intermediate at m/z 151 was observed in the photodegradation process of BPA by liquid chromatography combined with mass spectrometry (LC-MS) analysis, and a simple and hole-predominated photodegradation pathway of BPA was subsequently proposed. Furthermore, this photocatalyst

Heat Transfer Analysis of Methane Hydrate Sediment Dissociation in a Closed Reactor by a Thermal Method

Directory of Open Access Journals (Sweden)

Mingjun Yang

2012-05-01

Full Text Available The heat transfer analysis of hydrate-bearing sediment involved phase changes is one of the key requirements of gas hydrate exploitation techniques. In this paper, experiments were conducted to examine the heat transfer performance during hydrate formation and dissociation by a thermal method using a 5L volume reactor. This study simulated porous media by using glass beads of uniform size. Sixteen platinum resistance thermometers were placed in different position in the reactor to monitor the temperature differences of the hydrate in porous media. The influence of production temperature on the production time was also investigated. Experimental results show that there is a delay when hydrate decomposed in the radial direction and there are three stages in the dissociation period which is influenced by the rate of hydrate dissociation and the heat flow of the reactor. A significant temperature difference along the radial direction of the reactor was obtained when the hydrate dissociates and this phenomenon could be enhanced by raising the production temperature. In addition, hydrate dissociates homogeneously and the temperature difference is much smaller than the other conditions when the production temperature is around the 10 °C. With the increase of the production temperature, the maximum of ΔT_oi grows until the temperature reaches 40 °C. The period of ΔT_oi have a close relation with the total time of hydrate dissociation. Especially, the period of ΔT_oi with production temperature of 10 °C is twice as much as that at other temperatures. Under these experimental conditions, the heat is mainly transferred by conduction from the dissociated zone to the dissociating zone and the production temperature has little effect on the convection of the water in the porous media.

Fabrication, characterization and photocatalytic properties of Ag/AgI/BiOI heteronanostructures supported on rectorite via a cation-exchange method

Energy Technology Data Exchange (ETDEWEB)

Chen, Yunfang [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Fang, Jianzhang, E-mail: fangjzh@scnu.edu.cn [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Guangdong Technology Research Center for Ecological Management and Remediation of Urban Water System, Guangzhou 510006 (China); Lu, Shaoyou [Shenzhen Center for Disease Control and Prevention, Shenzhen 518055 (China); Wu, Yan; Chen, Dazhi; Huang, Liyan [Institute of Engineering Technology of Guangdong Province, Key Laboratory of Water Environmental Pollution Control of Guangdong Province, Guangzhou 510440 (China); Xu, Weicheng; Zhu, Ximiao [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Fang, Zhanqiang [School of Chemistry and Environment, South China Normal University, Guangzhou 510006, Guangdong (China); Guangdong Technology Research Center for Ecological Management and Remediation of Urban Water System, Guangzhou 510006 (China)

2015-04-15

Highlights: • Ag/AgI/BiOI-rectorite was prepared by twice cation-exchange process. • Ag/AgI/BiOI-rectorite photocatalyst possessed SPR and adsorption capacity. • Ag/AgI/BiOI-rectorite exhibited highly photocatalytic activity. • Trapped holes and ·O{sub 2}{sup −} were formed active species in the photocatalytic system. - Abstract: In this work, a new plasmonic photocatalyst Ag/AgI/BiOI-rectorite was prepared via a cation exchange process. The photocatalyst had been characterized by X-ray powder diffraction (XRD), Raman spectra, nitrogen sorption (BET), field-emission scanning electron microscope (FE-SEM), X-ray photoelectron spectroscopy (XPS) and UV–vis diffuse reflectance spectroscopy (DRS). The photocatalytic activity, which was evaluated by degradation of rhodamine B (RhB) and bisphenol A (BPA) under visible light irradiation, was enhanced significantly by loading Ag/AgI/BiOI nanoparticles onto rectorite. The photogenerated holes and superoxide radical (·O{sub 2}{sup −}) were both formed as active species for the photocatalytic reactions under visible light irradiation. The existence of metallic Ag particles, which possess the surface plasmon resonance effect, acted as an indispensable role in the photocatalytic reaction.

JENDL-4.0 benchmarking for fission reactor applications

International Nuclear Information System (INIS)

Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

2011-01-01

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

François Dosse

Directory of Open Access Journals (Sweden)

Marieta de Moraes Ferreira

2012-12-01

Full Text Available François Dosse, historiador francês, nasceu em Paris numa família de classe média, e desde cedo se interessou por política, vinculando-se quando jovem ao trotskismo. Estudou sociologia e história na Université de Vincennes - Paris VIII. Aprovado no exame de Agrégation, lecionou vários anos nos liceus de Pontoise e Boulogne-Billancourt. Foi Maître de conférences no IUFM (Instituto de Formação de Mestres de Versailles e no de Nanterre. Foi aprovado no exame para dirigir pesquisas em 2001, quando produziu um trabalho sobre Michel de Certeau e consolidou sua orientação para a área de teoria da história e historiografia. A convite de Henri Rousseau, vinculou-se ao IHTP, onde participou de vários seminários voltados à epistemologia dos estudos sobre o tempo presente. Publicou inúmeros trabalhos nessa área, focalizando especialmente biografias de intelectuais como Paul Ricoeur e Pierre Nora. Atualmente é Professor no IUFM de Créteil.

Works of shifting discharge facilities in construction for adding No.3 and No.4 plants to Oi Nuclear Power Station

International Nuclear Information System (INIS)

Matsuoka, Gen-ichi; Yoshida, Atsumu.

1989-01-01

At present in Oi Power Station, No.1 and No.2 plants of 1175 MWe output each are in operation, but in order to stabilize electric power supply for a long period, Kansai Electric Power Co., Inc. earnestly advances the construction works for adding No.3 and No.4 plants of each 1180 MWe output PWR. No.3 plant is expected to begin the operation in October, 1991, and No.4 plant in August, 1992. The works for creating the site were started in July, 1985, and the flat land of about 60,000 m 2 and the reclaimed land of about 80,000 m 2 were prepared. Subsequently, the main construction works were started in May, 1987, and the rate of general progress was 21 % in No.3 plant and 2 % in No.4 plant as of the end of October, 1988. Due to the addition of No.3 and No.4 plants, the quantity of condenser cooling water discharge increases to 318 m 3 /s from 150 m 3 /s at present, therefore, the bank having discharge holes is shifted from the present position about 100 m toward sea. As to the problems, the shifting works in flowing water, the method of shifting, the examination on lifting caissons and culverts, the trial construction of chemical anchors and so on were investigated. The execution of the shifting works is reported. (K.I.)

In situ grown hierarchical 50%BiOCl/BiOI hollow flowerlike microspheres on reduced graphene oxide nanosheets for enhanced visible-light photocatalytic degradation of rhodamine B

Science.gov (United States)

Su, Xiangde; Yang, Jinjin; Yu, Xiang; Zhu, Yi; Zhang, Yuanming

2018-03-01

50%BiOCl/BiOI/reduced graphene oxide (50%BiOCl/BiOI/rGO) composite photocatalyst was synthesized successfully by a facile one-step solvothermal route in this work. Reduction of graphene oxide (GO) took place in the process of solvothermal reaction and a new Bi-C bond between rGO and 50%BiOCl/BiOI was formed. The introduction of rGO affected the morphology of 50%BiOCl/BiOI, resulting in the transformation of 50%BiOCl/BiOI from solid microspheres to hollow microspheres. Both the introduction of rGO and formation of 50%BiOCl/BiOI hollow microspheres can facilitate the light absorption. The strong interaction between 50%BiOCl/BiOI and rGO and the electrical conductivity of rGO greatly improved the effective separation of photogenerated carriers. Hence, GOB-5 demonstrated the highest photocatalytic activity which was over twice of the pristine 50%BiOCl/BiOI in the presence of visible light. Mechanism study revealed that 50%BiOCl/BiOI generated electrons and holes in the presence of visible light, and holes together with rad O2- generated from reduction of O2 by electrons degraded the pollutant directly. Overall, this work provides an excellent reference to the synthesis of chemically bonded BiOX/BiOY (X, Y = Cl, Br, I)/rGO nanocomposite and helps to promote their applications in environmental protection and photoelectric conversion.

An ion exchange strategy to BiOI/CH{sub 3}COO(BiO) heterojunction with enhanced visible-light photocatalytic activity

Energy Technology Data Exchange (ETDEWEB)

Han, Qiaofeng, E-mail: hanqiaofeng@njust.edu.cn; Yang, Zhen; Wang, Li; Shen, Zichen; Wang, Xin; Zhu, Junwu; Jiang, Xiaohong

2017-05-01

Highlights: • BiOI/BiOAc heterojunction was firstly synthesized by an ion exchange route. • BiOI/BiOAc exhibited enhanced visible-light-driven photoreactivity for the dyes degradation in comparison with individuals. • Photocatalytic activity of the as-prepared BiOI/BiOAc is better than that prepared by precipitation-deposition method. • Photosensitization effect of BiOI to BiOAc was superior to that of Bi{sub 2}S{sub 3} due to suitable solubility constant. - Abstract: It is very significant to develop CH{sub 3}COO(BiO) (denoted as BiOAc) based photocatalysts for the removal of pollutants due to its non-toxicity and availability. We previously reported that BiOAc exhibited excellent photocatalytic activity for rhodamine B (RhB) degradation under UV light irradiation. Herein, by an ion exchange approach, BiOI/BiOAc heterojunction could be easily obtained. The as-prepared heterojunction possessed enhanced photodegradation activity for multiple dyes including RhB and methyl orange (MO) under visible light illumination in comparison with individual materials. Good visible-light photocatalytic activity of the heterojunction could be attributed to the increased visible light response, effective charge transfer from the modified band position and close interfacial contact due to partial ion exchange method.

Conférence | À quelle distance sommes-nous des Lumières ? | Etienne Klein et François-Xavier Verger | 4 mars

CERN Multimedia

2015-01-01

À quelle distance sommes-nous des Lumières ?, par Etienne Klein et François-Xavier Verger.   Mercredi 4 mars 2015 à 20h30 Globe de la science et de l'innovation Route de Meyrin, 1211 Genève Soirée en français Etienne Klein : L'esprit des Lumières promeut l’autonomie de l’entendement et la primauté de l’esprit scientifique. Il pose également la finalité humaine de nos actes – on ne vise pas Dieu, mais les hommes –, et l’universalité, notamment celle des droits de l’homme. Nous nous considérons volontiers comme les héritiers des Lumières. Mais à quelle distance sommes-nous réellement de leur esprit ? Etienne Klein est physicien, directeur de recherches au CEA et docteur en philosophie des sciences. Il est professeur &a...

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

InGaAs-OI Substrate Fabrication on a 300 mm Wafer

Directory of Open Access Journals (Sweden)

Sebastien Sollier

2016-09-01

Full Text Available In this work, we demonstrate for the first time a 300-mm indium–gallium–arsenic (InGaAs wafer on insulator (InGaAs-OI substrates by splitting in an InP sacrificial layer. A 30-nm-thick InGaAs layer was successfully transferred using low temperature direct wafer bonding (DWB and Smart CutTM technology. Three key process steps of the integration were therefore specifically developed and optimized. The first one was the epitaxial growing process, designed to reduce the surface roughness of the InGaAs film. Second, direct wafer bonding conditions were investigated and optimized to achieve non-defective bonding up to 600 °C. Finally, we adapted the splitting condition to detach the InGaAs layer according to epitaxial stack specifications. The paper presents the overall process flow that achieved InGaAs-OI, the required optimization, and the associated characterizations, namely atomic force microscopy (AFM, scanning acoustic microscopy (SAM, and HR-XRD, to insure the crystalline quality of the post transferred layer.

Jullien François, Le nu impossible

Directory of Open Access Journals (Sweden)

André Delobelle

2011-03-01

Full Text Available Bien qu’il ne s’agisse ici que d’esthétique, ce livret intéresse cependant la généralité des sciences humaines par la distinction extrêmement pertinente que l’A. introduit et développe entre les notions de forme externe et de forme interne. En partant du pro­blème posé par la représentation plastique du corps humain, François Jullien note en effet que l’artiste a toujours à choisir entre deux manières fondamentales : ou, comme dans les Idoles cycladiques, se limiter à une « géométrisation du ...

Observations of CO and OI in stars in globular clusters

International Nuclear Information System (INIS)

Wallerstein, G.; Pilachowski, C.

1978-01-01

Since studies at classification dispersion and early analyses of high dispersion spectra have yielded little quantitative data on the abundances of C, N, and O in globular clusters the authors have been endeavoring to establish their abundances in stars in several clusters. The problem has been approached in two ways, by observing the 2.3 micron CO bands and the 6300 A [OI] line in individual stars in globular clusters. (Auth.)

Chernobyl: recovery operations and the entombment of Reactor 4

International Nuclear Information System (INIS)

Dalziel, S.P.C.

1988-01-01

The immediate actions taken following the accident at the Chernobyl-number 4 reactor in April 1986 are described. These included actions to put out the fires, initial medical aid and the dropping of sand, lead, dolomite and boron onto the reactor from helicopters. Following this the chamber below the damaged reactor core was filled with concrete to prevent any further explosions or meltdown. The reactor was subsequently entombed in steel and concrete. The evacuation of the surrounding area is also mentioned. (U.K.)

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

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

Evaluation of stomatognathic problems in children with osteogenesis imperfecta (osteogenesis imperfecta - oi) - preliminary study.

Science.gov (United States)

Smoląg, Danuta; Kulesa-Mrowiecka, Małgorzata; Sułko, Jerzy

2017-01-01

According to epidemiological data, muscular dysfunctions of the masticatory system occur in 15-23% of the population. Preventive examinations of functional disorders of the stomatognathic system are, therefore, of particular importance. A distinct group of patients exposed to dysfunctions in the area of the masticatory organ locomotor apparatus comprises those with genetic diseases characterised by disorders in collagen formation. One of such diseases is osteogenesis imperfecta (OI) and dentinogenesis imperfecta that usually goes together with the former. The objective of this work was to evaluate the frequency with which particular disorders of the masticatory organ locomotor apparatus occur within the group of patients with osteogenesis imperfecta. The study was performed on patients of the Orthopaedic Clinic of the Polish-American Paediatric Institute in Kraków. The mean age of the children was 7.9 years. In all the cases, a genetic diagnosis of OI has been confirmed. The research methods were based on an in-depth interview on family diseases, pregnancy, postnatal period, feeding, subjective assessment of dysfunctions in the stomatognathic system. An examination of the deformations in the stomatognathic system and the skeleton was conducted, as well as an examination of the trauma and tone of the jaw. The relationship between breastfeeding and swallowing and speech disorders was also evaluated. The impact of intubation on mandibular ranges was investigated. The results obtained were subjected to statistical analysis on the basis of which conclusions were drawn concerning disorders in the stomatognathic system which tend to occur in children with OI. The renunciation of breastfeeding significantly contributes to sucking and swallowing disorders, rumen disorders, as well as biomechanical disorders in the temporomandibular joint. A significant dependence between breastfeeding and swallowing problems was found, whereas there was no such dependence with respect to

Evaluation of OiW Measurement Technologies for Deoiling Hydrocyclone Efficiency Estimation and Control

DEFF Research Database (Denmark)

Løhndorf, Petar Durdevic; Pedersen, Simon; Yang, Zhenyu

2016-01-01

Offshore oil and gas industry has been active in the North Sea for more than half a century, contributing to the economy and facilitating a low oil import rate in the producing countries. The peak production was reached in the early 2000s, and since then the oil production has been decreasing while...... to reach the desired oil production capacity, consequently the discharged amount of oil increases.This leads to oceanic pollution, which has been linked to various negative effects in the marine life. The current legislation requires a maximum oil discharge of 30 parts per million (PPM). The oil in water...... a novel control technology which is based on online and dynamic OiW measurements. This article evaluates some currently available on- line measuring technologies to measure OiW, and the possibility to use these techniques for hydrocyclone efficiency evaluation, model development and as a feedback...

Involving Families with Osteogenesis Imperfecta in Health Service Research: Joint Development of the OI/ECE Questionnaire.

Directory of Open Access Journals (Sweden)

Maman Joyce Dogba

Full Text Available Despite the growing interest in understanding the psycho-social impact of rare genetic diseases, few studies examine this concept and even fewer seek to obtain feedback from families who have lived the experience. The aim of this project was to involve families of children living with osteogenesis imperfecta (OI in the development of a tool to assess the impact of OI on the lives of patients and their families.This project used an integrated knowledge translation approach in which knowledge users (clinicians and people living with OI and their families were consulted throughout the four steps of development, that is: content mapping, item generation, tool appraisal and pre-testing of the questionnaires. The International Classification of Functioning and Health was used as a framework for content mapping. Based on a scoping review we selected two validated tools to use as a basis for developing the questionnaire. The final parent self-report version measured six domains: experience of diagnosis; use of health services; use of social and psychological support services; expectations about tertiary specialized centers; and socio-demographic information.A total of 27 out of 40 families receiving care at the Shriners Hospital for Children-Canada and invited to participate in the pre-test returned the completed questionnaires. In more than two-thirds of families (69%; n = 18 OI was suspected either at or within the first 3 months after birth. Up to 46% of families consulted between 3 and 5 doctors (46%; n = 12 prior to final diagnosis. The use of services by families varied from 0 to 16 consultations, 0 to 9 exploratory examinations and 1 to 10 types of allied health services. In the 12 months prior to the study, fewer than a quarter of children had been admitted, for treatment, for hospital stays of longer than 8 hours or to an emergency department (24% and 9% respectively. Only 29% of parents received psychological support.This joint development

François Valentijn Antara etika dan estetika

Directory of Open Access Journals (Sweden)

R.Z. Leirissa

2008-10-01

Full Text Available A controversial book on the history of the VOC in the seventeenth century waswritten by François Valentijn, a preacher in the payroll of the Dutch East IndiaCompany (VOC. The book (eight volumes, published in 1724-1726, was knownby its popular title Oud en Nieuw Oost-Indiën. Most of the materials in the bookwere pirated from other sources without acknowledging their authors as isthe proper practice of historiography. But as to its style, a number of Dutchwriters appreciated its esthetic qualities. Beside that, the book is indeed usefulto historians today because some of the materials pirated in the book have nowbeen lost forever.

Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

International Nuclear Information System (INIS)

Hu Jian; Jiang Nan; Li Jie; Shang Kefeng; Lu Na; Wu Yan; Mizuno Akira

2016-01-01

The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. (paper)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, Adimir dos; Nascimento, Jamil Alves do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

International Nuclear Information System (INIS)

Santos, A. dos; Nascimento, J.A. do

2002-01-01

An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

Enhanced Photoelectrocatalytic Activity of BiOI Nanoplate-Zinc Oxide Nanorod p-n Heterojunction.

Science.gov (United States)

Kuang, Pan-Yong; Ran, Jing-Run; Liu, Zhao-Qing; Wang, Hong-Juan; Li, Nan; Su, Yu-Zhi; Jin, Yong-Gang; Qiao, Shi-Zhang

2015-10-19

The development of highly efficient and robust photocatalysts has attracted great attention for solving the global energy crisis and environmental problems. Herein, we describe the synthesis of a p-n heterostructured photocatalyst, consisting of ZnO nanorod arrays (NRAs) decorated with BiOI nanoplates (NPs), by a facile solvothermal method. The product thus obtained shows high photoelectrochemical water splitting performance and enhanced photoelectrocatalytic activity for pollutant degradation under visible light irradiation. The p-type BiOI NPs, with a narrow band gap, not only act as a sensitizer to absorb visible light and promote electron transfer to the n-type ZnO NRAs, but also increase the contact area with organic pollutants. Meanwhile, ZnO NRAs provide a fast electron-transfer channel, thus resulting in efficient separation of photoinduced electron-hole pairs. Such a p-n heterojunction nanocomposite could serve as a novel and promising catalyst in energy and environmental applications. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

A study of the terrestrial thermosphere by remote sensing of OI dayglow in the far and extreme ultraviolet

International Nuclear Information System (INIS)

Cotton, D.M.

1991-01-01

The upper region of the Earth's atmosphere, the thermosphere, is a key part of the coupled solar-terrestrial system. An important method of obtaining information in the this region is through analysis of radiation excited through the interactions of the thermosphere with solar ionizing, extreme and far ultraviolet radiation. This dissertation presents one such study by the remote sensing of OI in the far and extreme ultraviolet dayglow. The research program included the development construction, and flight of a sounding rocket spectrometer to test this current understanding of the excitation and transport mechanisms of the OI 1356, 1304, 1027, and 989 angstrom emissions. This data set was analyzed using current electron and radiative transport models with the purpose of checking the viability of OI remote sensing; that is, whether existing models and input parameters are adequate to predict these detailed measurements. From discrepancies between modeled and measured emissions, inferences about these input parameters were made. Among other things, the data supports a 40% optically thick cascade contribution to the 1304 angstrom emission. From upper lying states corresponding to 1040, 1027 and 989 angstrom about half of this cascade has been accounted for in this study. There is also evidence that the Lyman β airglow from the geo-corona contributes a significant proportion (30-50%) to the OI 1027 angstrom feature. Furthermore, the photoelectron contribution to the 1027 angstrom feature appears to be underestimated in the current models by a factor of 20

Nuclear energy. The innovations of the N4 reactor

International Nuclear Information System (INIS)

Anon.

1998-01-01

The coupling to the electric network of the two first units of N4 type reactors, on the site of Chooz in the Ardennes, marks the third great step of the French nuclear programme of PWR type reactors, after the realization of 34 units of 900 MWe and 20 units of 1300 M We. The nuclear boiler N4, realizes a new evolution in power, in performances and in reliability. (N.C.)

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

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

International Nuclear Information System (INIS)

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

2007-01-01

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

François-Ronan Dubois, Introduction aux Porn Studies

OpenAIRE

Garnier, Guillaume

2014-01-01

Partant des débats qui entourent la pornographie et sa consommation (la pornographie est-elle sexiste ? Ou bien féministe ? Qu’en est-il de la moralité et de la législation ?…), François-Ronan Dubois, agrégé de lettres modernes et doctorant en littérature française, s’interroge sur les porn studies discipline novatrice originaire des États-Unis, qui s’est récemment imposée comme essentielle pour apporter un discours compétent sur la pornographie. Les porn studies s’emparant des formes et des ...

El concepto de límite en el B-Mu de François Roche

Directory of Open Access Journals (Sweden)

Eugenio Pandolfini

2012-12-01

Full Text Available

Resumen

Algunos proyectos, como el Dusty relief/B‐mu (2002 de François Roche demuestran como edificios complejos, que toman distancia desde los modelos mecanicistas para referirse a nuevos paradigmas, se pueden interpretar y comprenderse  mejor gracias a un análisis perceptivo que acerca el proyecto de arquitectura a cuestiones como la relación psicológica del hombre con la arquitectura, el miedo al espacio, y las  patologías  vinculadas a la percepción y a las neurosis modernas.
En este caso, aparte de las repercusiones que la fachada de  polvo podría tener en el ámbito de la ecología urbana, es interesante analizar algunos aspectos ligados a la dicotomía  entre forma externa y volúmenes internos para la que François Roche cita como referencia el raumplan de Adolf Loos, pero que presenta motivos para una reflexión vinculada a los aspectos  perceptivos.
El artículo trata de analizar como François Roche proyecta sus edificios extremando la dicotomía entre interior/visual y  exterior/táctil, desarrollando así una nueva relación con el lugar. Roche diseña la fachada exterior del B‐mu autoimponiéndose  una limitación del sentido de la vista, a favor de una dimensión háptica del proyecto y lo hace envolviendo los ámbitos arquitectónicos más familiares de una interfaz abstracta y táctil.

Palabras clave

edificio, percepción, entorno, envolvente, límite

Abstract

Some projects such as Dusty relief/B‐mu (2002 by François  Roche demonstrate how complex buildings, which distance themselves from the mechanicist models in order to refer  themselves to new paradigms, can be better understood and interpreted thanks to a perceptive analysis.
This analysis brings the architectural project closer to matters  such as man’s psychological relationship to Architecture, the fear of space

Notable light-free catalytic activity for pollutant destruction over flower-like BiOI microspheres by a dual-reaction-center Fenton-like process.

Science.gov (United States)

Wang, Liang; Yan, Dengbiao; Lyu, Lai; Hu, Chun; Jiang, Ning; Zhang, Lili

2018-10-01

BiOI is widely used as photocatalysts for pollutant removal, water splitting, CO 2 reduction and organic transformation due to its excellent photoelectric properties. Here, we report for the first time that a light-free catalyst consisting of the flower-like BiOI microspheres (f-BiOI MSs) exposing (1 0 1) and (1 1 0) crystal planes prepared by a hydrothermal method in ethylene glycol environment can rapidly eliminate the refractory BPA within only ∼3 min through a Fenton-like process. The reaction activity is ∼190 times higher than that of the conventional Fenton catalyst Fe 2 O 3 . A series of characterizations and experiments reveal the formation of the dual reaction centers on f-BiOI MSs. The electron-rich O centers efficiently reduce H 2 O 2 to OH, while the electron-poor oxygen vacancies capture electrons from the adsorbed pollutants and divert them to the electron-rich area during the Fenton-like reactions. By these processes, pollutants are degraded and mineralized quickly in a wide pH range. Our findings address the problems of the classical Fenton reaction and are useful for the development of efficient Fenton-like catalysts through constructing dual reaction centers. Copyright © 2018 Elsevier Inc. All rights reserved.

A basic design of SR4 instrumentation and control system for research reactor

International Nuclear Information System (INIS)

Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

2010-01-01

An SR4 instrumentation and control systems of research reactor is the equipment of nuclear research reactors as power protection devices and control systems. The equipment is to monitor safety parameters and process parameters in the state of reactor shut down, start-up, and in operation at fixed power. In the engineering of Instrumentation and control systems SR4 research reactor, its basic design consists of technical specifications of the reactor protection system devices, technical specifications of the reactor power control system devices, technical specifications information system devices, and systems process termination cabling as a support system. This basic design is used as the basis for the preparation of detailed design and subsequent engineering development of instrumentation systems and control system integrated. (author)

Reactor power cutback system test experience at YGN 4

International Nuclear Information System (INIS)

Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

1995-01-01

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

An adaptive hybrid EnKF-OI scheme for efficient state-parameter estimation of reactive contaminant transport models

KAUST Repository

El Gharamti, Mohamad; Valstar, Johan R.; Hoteit, Ibrahim

2014-01-01

Reactive contaminant transport models are used by hydrologists to simulate and study the migration and fate of industrial waste in subsurface aquifers. Accurate transport modeling of such waste requires clear understanding of the system's parameters, such as sorption and biodegradation. In this study, we present an efficient sequential data assimilation scheme that computes accurate estimates of aquifer contamination and spatially variable sorption coefficients. This assimilation scheme is based on a hybrid formulation of the ensemble Kalman filter (EnKF) and optimal interpolation (OI) in which solute concentration measurements are assimilated via a recursive dual estimation of sorption coefficients and contaminant state variables. This hybrid EnKF-OI scheme is used to mitigate background covariance limitations due to ensemble under-sampling and neglected model errors. Numerical experiments are conducted with a two-dimensional synthetic aquifer in which cobalt-60, a radioactive contaminant, is leached in a saturated heterogeneous clayey sandstone zone. Assimilation experiments are investigated under different settings and sources of model and observational errors. Simulation results demonstrate that the proposed hybrid EnKF-OI scheme successfully recovers both the contaminant and the sorption rate and reduces their uncertainties. Sensitivity analyses also suggest that the adaptive hybrid scheme remains effective with small ensembles, allowing to reduce the ensemble size by up to 80% with respect to the standard EnKF scheme. © 2014 Elsevier Ltd.

An adaptive hybrid EnKF-OI scheme for efficient state-parameter estimation of reactive contaminant transport models

KAUST Repository

El Gharamti, Mohamad

2014-09-01

Reactive contaminant transport models are used by hydrologists to simulate and study the migration and fate of industrial waste in subsurface aquifers. Accurate transport modeling of such waste requires clear understanding of the system\\'s parameters, such as sorption and biodegradation. In this study, we present an efficient sequential data assimilation scheme that computes accurate estimates of aquifer contamination and spatially variable sorption coefficients. This assimilation scheme is based on a hybrid formulation of the ensemble Kalman filter (EnKF) and optimal interpolation (OI) in which solute concentration measurements are assimilated via a recursive dual estimation of sorption coefficients and contaminant state variables. This hybrid EnKF-OI scheme is used to mitigate background covariance limitations due to ensemble under-sampling and neglected model errors. Numerical experiments are conducted with a two-dimensional synthetic aquifer in which cobalt-60, a radioactive contaminant, is leached in a saturated heterogeneous clayey sandstone zone. Assimilation experiments are investigated under different settings and sources of model and observational errors. Simulation results demonstrate that the proposed hybrid EnKF-OI scheme successfully recovers both the contaminant and the sorption rate and reduces their uncertainties. Sensitivity analyses also suggest that the adaptive hybrid scheme remains effective with small ensembles, allowing to reduce the ensemble size by up to 80% with respect to the standard EnKF scheme. © 2014 Elsevier Ltd.

Thermal and fast reactor benchmark testing of ENDF/B-6.4

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

Science.gov (United States)

Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

2016-03-01

The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

Application of WIMSD-4 for ''MARIA'' reactor lattice calculations

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

1993-12-01

A general description of the WIMSD-4 lattice code is given with the emphasis on available geometrical models. The difficulties encountered while modelling reactor lattices with the tubular type fuel elements are explained. Then the analysis of code options allowing to overcome these difficulties is carried out. Eventually, recommendations of options and input parameters for calculations of MARIA reactor lattice with satisfactory accuracy are given. During the work a set of modifications had to be introduced leading to a new code version called WIMS-S. Another version, under the name WIMS-T has been developed to allow for burnup calculations of the MARIA reactor lattice with improved resonance approach. (author). 14 refs, 6 figs, 10 tabs

CD4 lymphocyte counts and serum p24 antigen of no diagnostic value in monitoring HIV-infected patients with pulmonary symptoms

DEFF Research Database (Denmark)

Orholm, M; Nielsen, T L; Nielsen, Jens Ole

1990-01-01

The diagnostic value of the CD4 cell counts and the HIV p24 antigen were evaluated in a consecutive series of 105 HIV-infected patients experiencing 128 episodes of pulmonary symptoms which required bronchoscopy. One-third of patients with opportunistic infection (OI) had CD4 counts greater than ....... In conclusion, the CD4 cell counts and the presence of p24 antigen in serum had a very limited predictive value for the presence of OI in HIV-infected patients with pulmonary symptoms.......The diagnostic value of the CD4 cell counts and the HIV p24 antigen were evaluated in a consecutive series of 105 HIV-infected patients experiencing 128 episodes of pulmonary symptoms which required bronchoscopy. One-third of patients with opportunistic infection (OI) had CD4 counts greater than 0.......200 x 10(9)/l, and 60% of patients without OI had CD4 counts less than 0.200 x 10(9)/l; 47 and 42% of patients with and without OI, respectively, had detectable p24 antigen in serum. Only 36% of the patients with OI presented the combination of CD4 cells less than 0.200 x 10(9)/l and p24 in serum...

1300MVA steam-turbine generators for Kansai Electric Power's Oi Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

Oishi, N; Amagasa, N; Ito, H; Yagi, K [Mitsubishi Electric Corp., Kobe (Japan). Kobe Works

1977-06-01

Mitsubishi Electric has completed two 1300 MVA generators, equipped with 5500kW brushless exciters, that will be the No. 1 and No. 2 generators of the Oi plant. They are among the largest anywhere, and incorporate such technological innovations as water cooling of the stator coil and asymmetrical arrangement of the rotor slots. The article discusses generator specifications and construction, the brushless exciter, and the results of factory tests.

On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

International Nuclear Information System (INIS)

Platonov, P.A.

2007-01-01

Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru

Biodegradation of Jet Fuel-4 (JP-4) in Sequencing Batch Reactors

Science.gov (United States)

1992-06-01

nalw~eo %CUMENTATION PAGE__ _ _ _ _ _ _ _ _O 74S Ab -A258 020 L AW POi~W6 DATI .~ TYP AIMqm ,-& 0 U. glbs A~ I ma"&LFUN Mu BIODEGRADATION OF JET FUEL...Specific Objectives of This Proposal Are: 1. To assess the ability of C. resinae , P. chrysosporium and selected bacterial consortia to degrade individual...chemical components of JP-4. 2. To develop a sequencing batch reactor that utilizes C. resinae to degrade chemical components of JP-4 in contaminated

Experiment operations plan for the MT-4 experiment in the NRU reactor

International Nuclear Information System (INIS)

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

1983-06-01

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700 0 F)

Fast breeder reactors--lecture 4

International Nuclear Information System (INIS)

Marshall, W.; Davies, L.M.

1986-01-01

This paper discusses the economics of fast breeder reactors. An algebraic background is presented which represents the various views expressed by different nations regarding the cost of fast breeder reactors and their associated fuel cycle services, the timescale by which they might be available, and the simultaneous variations in the price of uranium. Actual presentations made by individual countries in recent discussions serve to verify the general nature of this present discussion. It is assumed that if nuclear power is to make a long term contribution to the needs of the world, the introduction of fast breeder reactors is both essential and necessary

In-reactor oxidation of zircaloy-4 under low water vapor pressures

Science.gov (United States)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

In-reactor oxidation of zircaloy-4 under low water vapor pressures

International Nuclear Information System (INIS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

Operating reactors licensing actions summary. Vol.4, No. 4

International Nuclear Information System (INIS)

1984-06-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

Tidal and solar cycle effects on the OI 5577 A, NaD and OH(8,3) airglow emissions observed at 23 deg S

International Nuclear Information System (INIS)

Takahashi, H.; Sahai, Y.; Batista, P.P.

1984-01-01

The upper mesosphere airglow emissions OI 5577, NaD and OH have been observed at Cachoeira Paulista (22.7 deg S; 45.0 deg W) Brazil. Nocturnal variations and their seasonal dependencies in amplitude and phase, and the annual variations of these emissions are presented, analysing the data obtained from 1977 to 1982 during the ascending phase of the last solar cycle. The nocturnal variations of the OI 5577 emission and the OH rotational temperature showed a significant semidiurnal oscillation, with the phase of maximum moving from midnight in January to early morning in June. Semiannual variation of the OI 5577 and NaD emissions with the maximum intensities in April/May and October/November were observed. The OH rotational temperature, however, showed an annual variation, maximum in summer and minimum in winter, while no significant seasonal variation was found in the OH emission intensities. Long-term intensity variations are also presented with the solar sunspot numbers and the 10.7 cm flux. (author)

L’Afrique et l’histoire des techniques. Hommage à François Sigaut Africa and the history of techniques. Tribute to François Sigaut

Directory of Open Access Journals (Sweden)

Monique Chastanet

2013-02-01

Full Text Available Spécialiste de l’histoire et de l’anthropologie des techniques des sociétés préindustrielles, dans le domaine de l’agriculture et de l’alimentation, François Sigaut (10 novembre 1940 – 2 novembre 2012 nous a quittés récemment, emporté en deux mois par un cancer, alors qu’il avait encore de nombreux travaux en chantier. Ses recherches étaient centrées sur l’Europe, mais il s’intéressait également aux autres continents, à l’Afrique en particulier. Cet intérêt était lié à sa démarche comparativ...

THE START UP STRATEGIES FOR THE BRAZILIAN MOBILE MARKET : OI'S CASE STUDY

OpenAIRE

ROBERTA FRANCO TERZIANI

2004-01-01

Esta dissertação tem como objetivo analisar as estratégias de marketing implementadas pela Oi, a terceira entrante no mercado brasileiro de telefonia móvel, desde o seu pré-lançamento comercial até junho de 2004 para identificar os aspectos que tiveram influência nos resultados alcançados na penetração do mercado da telefonia móvel. Apesar da forte retração da atividade econômica observada no Brasil nos últimos anos, particularmente de 2002 a meados de 2004,...

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

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

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

International Nuclear Information System (INIS)

DeHart, M.D.

1995-03-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

Questions about the reactor accident with Chernobyl-4

International Nuclear Information System (INIS)

Heijboer, R.J.

1986-01-01

The author presents an inventory of existing information about the Chernobyl-4 accident. Several possible scenarios are described and a comparison is drawn with the Three Mile Island-2 accident. The author concludes that the event is connected to an inherent instability of the RBMK-1000 reactor type. (G.J.P.)

Application of the neutron noise technique for measurement of reactivity for subcritical reactor RA-4

International Nuclear Information System (INIS)

Orso, J; Marenzana, A

2012-01-01

Reactor core RA-4 is divided into two parts that come together to start reactor. The reactor with core separate has the largest subcritical condition, this condition is more secure and therefore the reactor shutdown. In this paper measurements are made of the decay constant of the neutron prompt ' P ', using the α-Rossi and α-Feynman methods to calculate the reactivity of the reactor core for different positions. Both techniques are compared and reactivity is obtained for several position of the reactor core using the α-Rossi technical which is obtained a function that gives the reactivity depending on the separation of the core length. Both techniques are verified using a no multiplicative system. Reactivity values for different position of the core obtained by α-Rossi technique are: $[0 cm] = (-11+/-1) dollar, $[3 cm] = (-7+/-1) dollar, $[3.5 cm] (-5.5+/-0.8) dollar, $[4.2 cm] = (-3.8+/-0.3) dollar y $[4.5] = (-3.0+/-0.1) dollar (author)

François Couperin’s Allemandes for Harpsichord: Linguistic Characteristics of 18th Century French Oratory as an Interpretive Resource

Directory of Open Access Journals (Sweden)

Beatriz Pavan

2016-11-01

Full Text Available This text discusses the linguistic characteristics of French oratory of the eighteenth century as an interpretive resource for François Couperin’s allemandes for harpsichord. The main objective was to investigate the oratory and its equivalence with the musical discourse of this time and place as a tool for interpreting the genre. The methodological approach followed was an analysis of the linguistic variations of French oratory of the eighteenth century and its correspondence with the allemandes of François Couperin contained in the four Livres de pièces de clavecin. Upon establishing the close relationship between the spoken language and musical expression, it was possible to apply the study of the suggested discourse to interpretative decisions.

OI-57, a Genomic Island of Escherichia coli O157, Is Present in Other Seropathotypes of Shiga Toxin-Producing E. coli Associated with Severe Human Disease▿

Science.gov (United States)

Imamovic, Lejla; Tozzoli, Rosangela; Michelacci, Valeria; Minelli, Fabio; Marziano, Maria Luisa; Caprioli, Alfredo; Morabito, Stefano

2010-01-01

Strains of Shiga toxin-producing Escherichia coli (STEC) are a heterogeneous E. coli group that may cause severe disease in humans. STEC have been categorized into seropathotypes (SPTs) based on their phenotypic and molecular characteristics and the clinical features of the associated diseases. SPTs range from A to E, according to a decreasing rank of pathogenicity. To define the virulence gene asset (“virulome”) characterizing the highly pathogenic SPTs, we used microarray hybridization to compare the whole genomes of STEC belonging to SPTs B, C, and D with that of STEC O157 (SPT A). The presence of the open reading frames (ORFs) associated with SPTs A and B was subsequently investigated by PCR in a larger panel of STEC and in other E. coli strains. A genomic island termed OI-57 was present in SPTs A and B but not in the other SPTs. OI-57 harbors the putative virulence gene adfO, encoding a factor enhancing the adhesivity of STEC O157, and ckf, encoding a putative killing factor for the bacterial cell. PCR analyses showed that OI-57 was present in its entirety in the majority of the STEC genomes examined, indicating that it represents a stable acquisition of the positive clonal lineages. OI-57 was also present in a high proportion of the human enteropathogenic E. coli genomes assayed, suggesting that it could be involved in the attaching-and-effacing colonization of the intestinal mucosa. In conclusion, OI-57 appears to be part of the virulome of pathogenic STEC and further studies are needed to elucidate its role in the pathogenesis of STEC infections. PMID:20823207

Spray pyrolysis deposition and photoelectrochemical properties of n-type BiOI nanoplatelet thin films.

Science.gov (United States)

Hahn, Nathan T; Hoang, Son; Self, Jeffrey L; Mullins, C Buddie

2012-09-25

Bismuth oxy-iodide is a potentially interesting visible-light-active photocatalyst; yet there is little research regarding its photoelectrochemical properties. Herein we report the synthesis of BiOI nanoplatelet photoelectrodes by spray pyrolysis on fluorine-doped tin oxide substrates at various temperatures. The films exhibited n-type conductivity, most likely due to the presence of anion vacancies, and optimized films possessed incident photon conversion efficiencies of over 20% in the visible range for the oxidation of I(-) to I(3)(-) at 0.4 V vs Ag/AgCl in acetonitrile. Visible-light photons (λ > 420 nm) contributed approximately 75% of the overall photocurrent under AM1.5G illumination, illustrating their usefulness under solar light illumination. A deposition temperature of 260 °C was found to result in the best performance due to the balance of morphology, crystallinity, impurity levels, and optical absorption, leading to photocurrents of roughly 0.9 mA/cm(2) at 0.4 V vs Ag/AgCl. Although the films performed stably in acetonitrile, their performance decreased significantly upon extended exposure to water, which was apparently caused by a loss of surface iodine and subsequent formation of an insulating bismuth hydroxide layer.

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

International Nuclear Information System (INIS)

1992-07-01

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

OSCAR-4 Code System Application to the SAFARI-1 Reactor

International Nuclear Information System (INIS)

Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

2008-01-01

The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

CD4 descriptions at various clinical HIV/AIDS stages with tuberculosis and non-tuberculosis opportunistic infections at dr. Zainoel Abidin hospital in Banda Aceh, Indonesia

Science.gov (United States)

Shaleh, A. S.; Fahrial; Siregar, M. L.; Jamil, K. F.

2018-03-01

Human Immunodeficiency Virus (HIV) / Acquired Immune Deficiency Syndrome (AIDS) is a retrovirus that infects the human immune system. Damaged immunity in HIV/AIDS patients is marked by a decline in CD4 cells. Opportunistic infections appear differently, depending on the levels of immunosuppressive degrees, and the frequency of opportunistic infections in the environment. This study aims to describe the CD4 at various clinical stages of HIV/AIDS with tuberculosis and non-tuberculosis opportunistic infections (OI). Descriptive study with a cross-sectional design. This study is secondary data obtained from the medical records of patients with HIV/AIDS in the period of January 2011 - December 2015 at dr. Zainoel Abidin Hospital in Banda Aceh. The samples that met the inclusion criteria were 135 people with 63 cases were tuberculosis OI, 33 cases were non-tuberculosis and 39 cases were without OI. The study showed CD4 tuberculosis OI and 93.9% for non-tuberculosis OI.

Polyethylene imine-grafted ACF@BiOI{sub 0.5}Cl{sub 0.5} as a recyclable photocatalyst for high-efficient dye removal by adsorption-combined degradation

Energy Technology Data Exchange (ETDEWEB)

Li, Hongyan [Collaborative Innovation Center of Suzhou Nano Science and Technology, College of Chemistry, Chemical Engineering and Materials Science, Suzhou University, Suzhou, Jiangsu 215123 (China); Li, Najun, E-mail: linajun@suda.edu.cn [Collaborative Innovation Center of Suzhou Nano Science and Technology, College of Chemistry, Chemical Engineering and Materials Science, Suzhou University, Suzhou, Jiangsu 215123 (China); State and Local Joint Engineering Laboratory for Novel Functional Polymeric Materials, Suzhou, Jiangsu 215123 (China); Chen, Dongyun; Xu, Qingfeng [Collaborative Innovation Center of Suzhou Nano Science and Technology, College of Chemistry, Chemical Engineering and Materials Science, Suzhou University, Suzhou, Jiangsu 215123 (China); State and Local Joint Engineering Laboratory for Novel Functional Polymeric Materials, Suzhou, Jiangsu 215123 (China); Lu, Jianmei, E-mail: lujm@suda.edu.cn [Collaborative Innovation Center of Suzhou Nano Science and Technology, College of Chemistry, Chemical Engineering and Materials Science, Suzhou University, Suzhou, Jiangsu 215123 (China); State and Local Joint Engineering Laboratory for Novel Functional Polymeric Materials, Suzhou, Jiangsu 215123 (China)

2017-05-01

Highlights: • A recyclable photocatalyst was facilely fabricated by immobilization and grafting. • Contribution from each component in the composite towards enhanced performance. • High removal efficiency was achieved under adsorption-combined degradation. • The composite photocatalyst can be easily separated from water for direct reuse. - Abstract: A recyclable photocatalyst with adsorption property was prepared for high-efficient complete removal of anionic dyes from water by synergetic adsorption and photocatalytic degradation. Firstly, binary bismuth oxyhalide composed as BiOI{sub 0.5}Cl{sub 0.5} was immobilized on activated carbon fibers (ACF) to get a recyclable photocatalyst (ACF@BiOI{sub 0.5}Cl{sub 0.5}) via one-step solvothermal method. Then it was modified with branched polyethylene imine (PEI) whose abundant amino groups can adsorb contaminants from water by electrostatic interaction. SEM images showed that the nanosheets-based flower-like photocatalytic microspheres uniformly distributed on the ACF surface after grafting of small amount of PEI. But from TGA results we can deduce that the percentage of PEI grafted onto ACF@BiOI{sub 0.5}Cl{sub 0.5} is about 18 wt%. During the synergistic process, the grafted PEI and immobilized BiOI{sub 0.5}Cl{sub 0.5} are worked as the adsorbent and the photocatalyst, respectively. In addition, ACF, as flexible, conductive and corrosion-resistant supports, are beneficial to the photocatalytic degradation process. So the obtained composite PEI-g-ACF@BiOI{sub 0.5}Cl{sub 0.5} has a high removal efficiency of contaminants under visible light irradiation with the synergistic effect of adsorption and photocatalytic degradation. And after facial separation without centrifuge, it can be reused without regeneration because of the real-time complete degradation of the adsorbed contaminants on the surface of the composite photocatalyst.

Effect of 4-methoxyindole-3-carbinol on the proliferation of colon cancer cells in vitro, when treated alone or in combination with indole-3-carbinol

DEFF Research Database (Denmark)

Kronbak, Remy; Duus, Fritz; Vang, Ole

2010-01-01

Consumption of cruciferous vegetables and cancer prevention seem to be positively associated. We present an easy two-step synthesis for 4-methoxyindole-3-carbinol (4MeOI3C), the expected breakdown product of 4-methoxyglucobrassicin during ingestion. 4MeOI3C inhibited the proliferation of human co...

Dos lienzos firmados por Alonso del Arco en la parisina iglesia Saint-François de Sales

Directory of Open Access Journals (Sweden)

Delenda, Odile

2004-09-01

Full Text Available El inventario detallado realizado en las iglesias de París por los conservadores des Oeuvres d'Art Religieuses et Civiles del ayuntamiento de la capital francesa, puede reservar gratas sorpresas. Muy recientemente Guénola Groud, conservateur du Patrimoine y su equipo, pudieron localizar en la «Salle des mariages» de la nueva iglesia de Saint-François-de-Sales (Paris, XVIIème dos importantes lienzos de medio punto, la Adoración de los Pastores (Fig. 2 y la Adoración de los Magos (Fig. 1 firmados por el pintor Alonso del Arco (1635-1704 colaborador y discípulo de Antonio Pereda. Existen en efecto dos iglesias Saint-François-de-Sales. La más antigua, de 1873, está situada en la calle Brémontier mientras que la más reciente, construida entre 1911 y 1913, tiene su entrada principal en la calle Ampère…

PODAAC-AQR40-4U7CS

Data.gov (United States)

National Aeronautics and Space Administration — The IPRC/SOEST Aquarius OI-SSS v4 product is a level 4, near-global, 0.5 degree spatial resolution, 7-day, optimally interpolated salinity dataset based on version...

SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

Summary of the 4th workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

Summary of the 4th workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

CFD for Nuclear Reactor Safety Applications (CFD4NRS-4) - Workshop Proceedings

International Nuclear Information System (INIS)

2014-01-01

Following the CFD4NRS workshops held in Garching, Germany (Sept. 2006), Grenoble, France (Sep. 2008) and Washington D.C., USA (Sept. 2010), this Workshop is intended to extend the forum created for numerical analysts and experimentalists to exchange information in the application of CFD and CMFD to NRS issues and in guiding nuclear reactor design thinking. The workshop includes single-phase and multi-phase CFD applications, and offers the opportunity to present new experimental data for CFD validation. More emphasis has been given to the experiments, especially on two-phase flow, for advanced CMFD modelling for which sophisticated measurement techniques are required. Understanding of the physics has been depen before starting numerical analysis. Single-phase and multi-phase CFD simulations with a focus on validation were performed in areas such as: single-phase heat transfer, boiling flows, free-surface flows, direct contact condensation and turbulent mixing. These relate to NRS-relevant issues, such as pressurised thermal shock, critical heat flux, pool heat exchangers, boron dilution, hydrogen distribution in containments, thermal striping, etc. The use of systematic error quantification and the application of BPGs were strongly encouraged. Experiments providing data suitable for CFD or CMFD validation were also presented. These included local measurements using multi-sensor probes, laser-based techniques (LDV, PIV or LIF), hot-film/wire anemometry, imaging, or other advanced measuring techniques. There were over 150 registered participants at the CFD4NRS-4 workshop. The programme consisted of 48 technical papers. Of these, 44 were presented orally and 4 as posters. An additional 8 posters related to the OECD/NEA-KAERI sponsored CFD benchmark exercise on turbulent mixing in a rod bundle with spacers (MATiS-H) were presented and a special session was allocated for 6 video presentations. In addition, five keynote lectures were given by distinguished experts. The

9 CFR 75.4 - Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic...

Science.gov (United States)

2010-01-01

... infectious anemia reactors and approval of laboratories, diagnostic facilities, and research facilities. 75.4... IN HORSES, ASSES, PONIES, MULES, AND ZEBRAS Equine Infectious Anemia (swamp Fever) § 75.4 Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic facilities, and...

Montag e a memória perdida:notas sobre Fahrenheit 451 de François Truffaut

OpenAIRE

Silva, Terezinha Elisabeth da

2003-01-01

The paper analyzes aspects related to books and memory, in François Truffault’s Fahrenheit 451. It discusses issues related to the prohibition and destruction of books by totalitarian regimes. It emphasizes Montag’s trajectory through the movie and his transformation into a book protector and point out characteristics of the society portrayed in the film, specially its fondness for the image and for the oral information.

Jean-François Gallotte, Joëlle Malberg, Carbone 14 le film, Les Mutins de Pangée

OpenAIRE

Curien, Julie

2014-01-01

Carbone 14 le film est un témoignage sur le mouvement des radios libres, à travers l'exemple de « Carbone 14 », « la radio active » devenue « la radio qui vous encule par les oreilles ». Filmé en 4 jours à la rentrée 1982, avec du matériel volé, le film ovni de Jean-François Galotte et Joëlle Malberg est sélectionné au Festival de Cannes en 1983. Il fait un tollé qui retombe comme une crêpe, aux oubliettes : on n'en parle plus avant... 2011, trente ans après la création de la radio culte et c...

Short- and Long-Timescale Thermospheric Variability as Observed from OI 630.0 nm Dayglow Emissions from Low Latitudes

Science.gov (United States)

Pallamraju, Duggirala; Das, Uma; Chakrabarti, Supriya

2011-01-01

We carried out high-cadence (5 min) and high-spatial resolution (2deg magnetic latitude) observations of daytime OI 630.0 nm airglow emission brightness from a low-latitude station to understand the behavior of neutral dynamics in the daytime. The results indicate that the wave periodicities of 12.20 min, and 2 h exist over a wide spatial range of around 8deg-12deg magnetic latitudes. The 20.80 min periodicities in the dayglow seem to appear more often in the measurements closer to the magnetic equator and not at latitudes farther away. Further, periodicities in that range are found to be frequent in the variations of the equatorial electrojet (EEJ) strength as well. We show that wave periodicities due to the neutral dynamics, at least until around 8deg magnetic latitude, are influenced by those that affect the EEJ strength variation as well. Furthermore, the average daily OI 630.0 nm emission brightness over 3 months varied in consonance with that of the sunspot numbers indicating a strong solar influence on the magnitudes of dayglow emissions.

A Dynamic Assessment of the François-Mitterrand Library

Directory of Open Access Journals (Sweden)

Valerie Vesque-Jeancard

2008-09-01

Full Text Available The François-Mitterrand Library, main site of the Bibliothèque nationale de France (BnF, was built in 1993 by architect Dominique Perrault. The library, which hosts 13 million documents, welcomes 3,500 visitors per day, either for consultation of documents in its majestic reading rooms or for conferences and exhibitions. The evolution of services provided to the library's visitors (increase in electronic resources for instance, the broadening of the BnF's users (from researchers to a broader public, including professionals and young people, the change in the building's environment (public transportation, urban extension, etc. along with the evolution of regulations (facilities for the disabled, etc. or, more recently, increased consideration for sustainable development, led to a continuous adaptation of the building throughout the last 15 years. This dynamic assessment will continue to be a key factor for the library to successfully face future challenges.

O amor cortês pelo avesso: François Villon e o debate sobre o Roman de la rose

Directory of Open Access Journals (Sweden)

Daniel Padilha Pacheco da Costa

2014-09-01

Full Text Available Resumo: Neste artigo, pretende-se discutir a paródia do amor cortês pelos lamentos burlescos do Testament de François Villon, com base nos preceitos e modelos que orientavam a invenção das letras na época. Complementares ao lamento do próprio testador pelo amor louco da sua juventude, os Regrets de la belle heaulmière utilizam como modelo poético o sermão da Velha do Roman de la rose. A imitação de uma das passagens desse poema mais duramente censuradas por Christine de Pisan evidencia que esses lamentos só podem ser compreendidos à luz do debate sobre o Roman de la rose, realizado no início do séc. XV na França. Dessa perspectiva, a paródia deve ser considerada não como uma recusa da tradição cortês no final da Idade Média, como pela crítica contemporânea, mas como um gênero particular da poesia burlesca visando a ridicularização do amor louco. Palavras-chave: François Villon; paródia; amor cortês; debate sobre o Roman de la rose; lamentos burlescos. Abstract: This paper intends to discuss the parody of courtly love performed by the burlesque regrets of François Villon’s Testament, using the poetic precepts and models based on which the writing was invented at the time. Complementary to the regret of the testator himself for the mad love of his youth, the Regrets de la belle heaulmière use as a poetic model the Old Woman’s sermon of the Romance of the rose. The imitation of one of the passages of this poem most harshly criticized by Christine de Pisan shows that those regrets can only be understood in the light of the debate of the Romance of the rose at the beginning of the XVth century in France. From this point of view, his parody must be considered not as a rejection of the courtly tradition in the late Middle Ages, as it is by contemporary criticism, but as a particular genre of burlesque poetry aiming to mock mad love. Keywords: François Villon; parody; courtly love; debate on the Roman de la rose

Present status of fusion reactor materials, 4

International Nuclear Information System (INIS)

Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

1982-01-01

Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

Operative management and outcomes in 103 AAST-OIS grades IV and V complex hepatic injuries: trauma surgeons still need to operate, but angioembolization helps.

Science.gov (United States)

Asensio, Juan A; Roldán, Gustavo; Petrone, Patrizio; Rojo, Esther; Tillou, Areti; Kuncir, Eric; Demetriades, Demetrios; Velmahos, George; Murray, James; Shoemaker, William C; Berne, Thomas V; Chan, Linda

2003-04-01

American Association for the Surgery of Trauma (AAST) Organ Injury Scale (OIS) grades IV and V complex hepatic injuries are highly lethal. Our objectives were to review experience and identify predictors of outcome and to evaluate the role of angioembolization in decreasing mortality. This was a retrospective 8-year study of all patients sustaining AAST-OIS grades IV and V hepatic injuries managed operatively. Statistical analysis was performed using univariate and multivariate logistic regression. The main outcome measure was survival. The study included 103 patients, with a mean Revised Trauma Score of 5.61 +/- 2.55 and a mean Injury Severity Score of 33 +/- 9.5. Mechanism of injury was penetrating in 80 (79%) and blunt in 23 (21%). Emergency department thoracotomy was performed in 21 (25%). AAST grade IV injuries occurred in 51 (47%) and grade V injuries occurred in 52 (53%). Mean estimated blood loss was 9,414 mL. Overall survival was 43%. Adjusted overall survival rate after emergency department thoracotomy patients were excluded was 58%. Results stratified to AAST-OIS injury grade were as follows: grade IV, 32 of 51 (63%); grade V, 12 of 52 (23%); grade IV versus grade V (p Trauma Score (adjusted p hepatic veins (adjusted p hepatic injuries.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

Science.gov (United States)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis

International Nuclear Information System (INIS)

Hoogenboom, J.E.

2013-01-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)

The z=0.0912 and z=0.2212 damped Ly alpha galaxies along the sight line toward the quasar OI 363

NARCIS (Netherlands)

Turnshek, DA; Rao, S; Nestor, D; Lane, W; Monier, E; Bergeron, J; Smette, A

2001-01-01

New optical and infrared observations along the sight line toward the quasar OI 363 (0738+313) are presented and discussed. Excluding quasars selectively observed because they were known to be located behind gas-rich galaxies and systems which lack confirming UV spectroscopic observations of the

Hyperplastic callus formation in osteogenesis imperfecta type V mimicking osteosarcoma: 4-year follow-up with resolution

Energy Technology Data Exchange (ETDEWEB)

Vieira, R.L.V.; Amaral, D.T. [Federal University of Sao Paulo, Department of Radiology, Sao Paulo (Brazil); Jesus-Garcia, Filho R. [Federal University of Sao Paulo, Department of Orthopedic Surgery, Sao Paulo (Brazil); Saraiva, G. [Federal University of Sao Paulo, Department of Endocrinology, Sao Paulo (Brazil); Fernandes, A.R.C. [University of California San Diego, Department of MSK Radiology, San Diego, CA (United States); Resnick, D.

2006-06-15

We report a case of hyperplastic callus formation that occurred in both femurs in a patient with type V osteogenesis imperfecta (OI), with 4-year follow-up and resolution. The clinical, histological and imaging aspects of this condition are discussed. Recognition of the hyperplastic callus formation in this particular type of OI is important in order to avoid misdiagnosis. (orig.)

Hyperplastic callus formation in osteogenesis imperfecta type V mimicking osteosarcoma: 4-year follow-up with resolution

International Nuclear Information System (INIS)

Vieira, R.L.V.; Amaral, D.T.; Jesus-Garcia, Filho R.; Saraiva, G.; Fernandes, A.R.C.; Resnick, D.

2006-01-01

We report a case of hyperplastic callus formation that occurred in both femurs in a patient with type V osteogenesis imperfecta (OI), with 4-year follow-up and resolution. The clinical, histological and imaging aspects of this condition are discussed. Recognition of the hyperplastic callus formation in this particular type of OI is important in order to avoid misdiagnosis. (orig.)

Characteristics of InAs/InGaAs/GaAs QDs on GeOI substrates with single-peak 1.3 µm room-temperature emission

International Nuclear Information System (INIS)

Liang, Y Y; Yoon, S F; Loke, W K; Ngo, C Y; Fitzgerald, E A

2012-01-01

GaAs-based quantum dot (QD) systems, especially InAs/InGaAs/GaAs QDs, have demonstrated superior device performances as compared with higher dimensional systems. However, to realize high-speed optical interconnects for Si-based electronics, one will need to grow the QDs on Si substrates. While it is promising to integrate the InAs/InGaAs/GaAs QDs on Si with the use of germanium-on-insulator-on-silicon (GeOI) substrates, reported results exhibit bimodal QD sizes and double emission peaks, i.e. unsatisfactory for realistic applications. In this paper, we showed that with an optimized GaAs buffer, single-peak 1.33 µm room-temperature emission can be obtained from InAs/InGaAs/GaAs QDs on GeOI substrates. (paper)

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

Calculations of fuel burn-up and radionuclide inventory in the syrian miniature neutron source reactor using the WIMSD4 code

International Nuclear Information System (INIS)

Khattab, K.

2005-01-01

Calculations of the fuel burn up and radionuclide inventory in the Miniature Neutron Source Reactor after 10 years (the reactor core expected life) of the reactor operating time are presented in this paper. The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnt up and plutonium produced in the reactor core, the concentrations and radioactivities of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well

Characteristics of Al Alloy as a Material for Hydrolysis Reactor of NaBH4

International Nuclear Information System (INIS)

Jung, Hyeon-Seong; Oh, Sung-June; Jeong, Jae-Jin; Na, Il-Chai; Chu, Cheun-Ho; Park, Kwon-Pil; Chu, Cheun-Ho

2015-01-01

Aluminum alloy was examined as a material of low weight reactor for hydrolysis of NaBH 4 . Aluminum is dissolved with alkali, but there is NaOH as a stabilizer in NaBH 4 solution. To decrease corrosion rate of aluminum, decrease NaOH concentration and this result in loss of NaBH 4 during storage of NaBH 4 solution. Therefore stability of NaBH 4 and corrosion of aluminum should be considered in determining the optimum NaOH concentration. NaBH 4 stability and corrosion rate of aluminum were measured by hydrogen evolution rate. NaBH 4 stability was tested at 20-50 .deg. C and aluminum corrosion was measured at 60-90 .deg. C. The optimum concentration of NaOH was 0.3 wt%, considering both NaBH 4 stability and aluminun corrosion. NaBH 4 hydrolysis reaction continued 200min in aluminum No 6061 alloy reactor with 0.3 wt% NaOH at 80-90 .deg. C.

Opportunistic Infections (OIs) in Patients with Hematologic Malignancies (HM) Treated with Bruton’s Tyrosine Kinase (BTK) and Phosphoinositide 3 Kinase (PI3K) Inhibitors: An 8-Year Retrospective Cohort Study

OpenAIRE

Issa, Nicolas; Arbona-Haddad, Esther; Nevett-Fernandez, Alexandra; Prestes, Daniel; Liakos, Alexis; Woolley, Ann; Hammond, Sarah; Brown, Jennifer; Baden, Lindsey; Marty, Francisco

2017-01-01

Abstract Background: BTK and PI3K inhibitors are increasingly used for treatment in patients with HM. OIs when these agents were used as first line therapy signaled an increased level of immunosuppression beyond what was expected from the mechanism of action of these drugs. The epidemiology of OIs in the setting of BTK and PI3K inhibitor use has not been characterized. Methods: We retrospectively studied a cohort of patients with HM who received BTK (ibrutinib, acalabrutinib, spebrutinib) or ...

Removal of FePO4 and Fe3(PO4)2 crystals on the surface of passive fillers in Fe0/GAC reactor using the acclimated bacteria

International Nuclear Information System (INIS)

Lai, Bo; Zhou, Yuexi; Yang, Ping; Wang, Juling; Yang, Jinghui; Li, Huiqiang

2012-01-01

Highlights: ► Fe 3 (PO 4 ) 2 and FePO 4 crystals would weaken treatment efficiency of Fe 0 /GAC reactor. ► Fe 3 (PO 4 ) 2 and FePO 4 crystals could be removed by the acclimated bacteria. ► FeS and sulfur in the passive film would be removed by the sulfur-oxidizing bacteria. ► Develop a cost-effective bio-regeneration technology for the passive fillers. - Abstract: As past studies presented, there is obvious defect that the fillers in the Fe 0 /GAC reactor begin to be passive after about 60 d continuous running, although the complicated, toxic and refractory ABS resin wastewater can be pretreated efficiently by the Fe 0 /GAC reactor. During the process, the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film are formed by the reaction between PO 4 3− and Fe 2+ /Fe 3+ . Meanwhile, they obstruct the formation of macroscopic galvanic cells between Fe 0 and GAC, which will lower the wastewater treatment efficiency of Fe 0 /GAC reactor. In this study, in order to remove the Fe 3 (PO 4 ) 2 and FePO 4 crystals on the surface of the passive fillers, the bacteria were acclimated in the passive Fe 0 /GAC reactor. According to the results, it can be concluded that the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film could be decomposed or removed by the joint action between the typical propionic acid type fermentation bacteria and sulfate reducing bacteria (SRB), whereas the PO 4 3− ions from the decomposition of the Fe 3 (PO 4 ) 2 and FePO 4 crystals were released into aqueous solution which would be discharged from the passive Fe 0 /GAC reactor. Furthermore, the remained FeS and sulfur (S) in the passive film also can be decomposed or removed easily by the oxidation of the sulfur-oxidizing bacteria. This study provides some theoretical references for the further study of a cost-effective bio-regeneration technology to solve the passive problems of the fillers in the zero-valent iron (ZVI) or Fe 0 /GAC reactor.

Flux distribution by neutrons semi-conductors detectors during the startup of the EL4 reactor

International Nuclear Information System (INIS)

Fuster, S.; Tarabella, A.

1967-01-01

The Cea developed neutron semi-conductors detectors which allows a quasi-instantaneous monitoring of neutrons flux distribution, when placed in a reactor during the tests. These detectors have been experimented in the EL4 reactor. The experiment and the results are presented and compared with reference mappings. (A.L.B.)

The different generation of nuclear reactors from Generation-1 to Generation-4

International Nuclear Information System (INIS)

Cognet, G.

2010-01-01

In this work author deals with the history of the development of nuclear reactors from Generation-1 to Generation-4. The fuel cycle and radioactive waste management as well as major accidents are presented, too.

Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

2017-01-01

At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

‘Sand’s Way’: The Voices of George Sand’s François the Waif in Marcel Proust’s Remembrance of Things Past

Directory of Open Access Journals (Sweden)

Françoise Grauby

2015-09-01

Full Text Available This paper traces one of the origins of Marcel Proust’s artistic vocation in his fascination for a novel by George Sand, François le Champi (François the Waif. In Remembrance of Things Past, the adult writer explores the gradual recognition of this early phase of his formation: Sand’s novel appears in the ‘Combray’ section in Swann’s Way and it reappears at the moment of apparent illumination regarding his future as a writer in Time Regained. Leaving deliberately aside the psychoanalytic implications of the story, this article will instead emphasise the ‘vocality’ of the story, taken from the oral tradition of the provincial Berry countryside and imbued with colloquial texture, to show how the retelling of this moment in time when the novel was encountered includes a definition of what artistic vocation is for the narrator.

The Influence of François Delsarte’s Ideas on the Modernity of Dance

Directory of Open Access Journals (Sweden)

Maria Albertina Silva Grebler

2012-11-01

Full Text Available This study discusses the influence François Delsarte’s ideas had on the pioneers of modern dance. Having rejected books on gestural rhetoric of his own time to conduct a rigorous study on human behavior in strong emotional situations, Delsarte established principles and systematized exercises that provided the pioneers of modern dance an alternative method of interpretation and creation, more centered on the subject and on their relationship with the environment, than on the imitation of traditional gestures. Delsarte’s system allowed modern creators to consider the took as a raw material for a research that was no longer defined by traditional techniques, but by the pursuit of a subjective form of corporeality.

CD4 lymphocyte counts and serum p24 antigen of no diagnostic value in monitoring HIV-infected patients with pulmonary symptoms

DEFF Research Database (Denmark)

Orholm, M; Nielsen, T L; Nielsen, Jens Ole

1990-01-01

The diagnostic value of the CD4 cell counts and the HIV p24 antigen were evaluated in a consecutive series of 105 HIV-infected patients experiencing 128 episodes of pulmonary symptoms which required bronchoscopy. One-third of patients with opportunistic infection (OI) had CD4 counts greater than 0....... In conclusion, the CD4 cell counts and the presence of p24 antigen in serum had a very limited predictive value for the presence of OI in HIV-infected patients with pulmonary symptoms....

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)

Safety of research reactors. Topical issues paper no. 4

International Nuclear Information System (INIS)

Alcala-Ruiz, F.; Ferraz-Bastos, J.L.; Kim, S.C.; Voth, M.; Boeck, H.; Dimeglio, F.; Litai, D.

2001-01-01

Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety review of the research reactor facility and to verify compliance with the IAEA Safety Standards. The methods used during an INSARR mission have been collected and analysed. Some of the important issues identified are the following: general ageing of the facility; uncertain status of many research reactors (in extended shutdown); indefinite deferral of return to operation or decommissioning; inadequate regulatory supervision; insufficient systematic (periodic) reassessment of safety; lack of quality assurance (QA) programmes; lack of an international safety convention or arrangement; lack of financial support for safety measures (e.g. safety reassessment, safety upgrading, decommissioning) and utilization; lack of clear utilization programmes; inadequate emergency preparedness; inadequate safety documentation (e.g. safety analysis report, operating rules and procedures, emergency plan); inadequate funding of shutdown reactors; weak safety culture; loss of expertise and corporate memory; loss of information concerning radioactive materials contained in retired experimental devices stored in the facility indefinitely; obsolescence of equipment and lack of spare parts; inadequate training and qualifications of regulators and operators; safety implications of new fuel types. These issues have been addressed by the IAEA Secretariat and the chairman of the International Nuclear Safety Advisory Group (INSAG). INSAG has identified three major safety issues that are: the increasing age of research reactors, the number of research reactors that are not operating anymore but have not been decommissioned, and the number of research reactors in countries that do not have appropriate regulatory authorities. This issue paper discusses the concerns generated by an analysis of the results of INSARR missions and those expressed by INSAG. The

[François de Lapeyronie, from Montpellier (1678-1747). "Surgery restorer" and universal spirit. The soul, Musc, rooster eggs].

Science.gov (United States)

Fischer, Louis-Paul; Ferrandis, Jean-Jacques; Blatteau, Jean-Eric

2009-01-01

François de Lapeyronie was a master in surgery in 1695 in Paris then in 1717 and rewarded with the rank of Medical Doctor of the University of Reims. The authors try to underline his intelligence and his broadmindedness through three publications about the centre of the soul in the corpus callosum, the anatomical dissection of a kind of stone marten and the scientific research of the so called 'egg of cock'.

Study of cold neutron sources: Implementation and validation of a complete computation scheme for research reactor using Monte Carlo codes TRIPOLI-4.4 and McStas

International Nuclear Information System (INIS)

Campioni, Guillaume; Mounier, Claude

2006-01-01

The main goal of the thesis about studies of cold neutrons sources (CNS) in research reactors was to create a complete set of tools to design efficiently CNS. The work raises the problem to run accurate simulations of experimental devices inside reactor reflector valid for parametric studies. On one hand, deterministic codes have reasonable computation times but introduce problems for geometrical description. On the other hand, Monte Carlo codes give the possibility to compute on precise geometry, but need computation times so important that parametric studies are impossible. To decrease this computation time, several developments were made in the Monte Carlo code TRIPOLI-4.4. An uncoupling technique is used to isolate a study zone in the complete reactor geometry. By recording boundary conditions (incoming flux), further simulations can be launched for parametric studies with a computation time reduced by a factor 60 (case of the cold neutron source of the Orphee reactor). The short response time allows to lead parametric studies using Monte Carlo code. Moreover, using biasing methods, the flux can be recorded on the surface of neutrons guides entries (low solid angle) with a further gain of running time. Finally, the implementation of a coupling module between TRIPOLI- 4.4 and the Monte Carlo code McStas for research in condensed matter field gives the possibility to obtain fluxes after transmission through neutrons guides, thus to have the neutron flux received by samples studied by scientists of condensed matter. This set of developments, involving TRIPOLI-4.4 and McStas, represent a complete computation scheme for research reactors: from nuclear core, where neutrons are created, to the exit of neutrons guides, on samples of matter. This complete calculation scheme is tested against ILL4 measurements of flux in cold neutron guides. (authors)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

International Nuclear Information System (INIS)

2007-03-01

In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

International Nuclear Information System (INIS)

2009-01-01

In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-02-15

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Characteristics of Al Alloy as a Material for Hydrolysis Reactor of NaBH{sub 4}

Energy Technology Data Exchange (ETDEWEB)

Jung, Hyeon-Seong; Oh, Sung-June; Jeong, Jae-Jin; Na, Il-Chai; Chu, Cheun-Ho; Park, Kwon-Pil [Sunchon National University, Suncheon (Korea, Republic of); Chu, Cheun-Ho [ETIS Co, Gimpo (Korea, Republic of)

2015-12-15

Aluminum alloy was examined as a material of low weight reactor for hydrolysis of NaBH{sub 4}. Aluminum is dissolved with alkali, but there is NaOH as a stabilizer in NaBH{sub 4} solution. To decrease corrosion rate of aluminum, decrease NaOH concentration and this result in loss of NaBH{sub 4} during storage of NaBH{sub 4} solution. Therefore stability of NaBH{sub 4} and corrosion of aluminum should be considered in determining the optimum NaOH concentration. NaBH{sub 4} stability and corrosion rate of aluminum were measured by hydrogen evolution rate. NaBH{sub 4} stability was tested at 20-50 .deg. C and aluminum corrosion was measured at 60-90 .deg. C. The optimum concentration of NaOH was 0.3 wt%, considering both NaBH{sub 4} stability and aluminun corrosion. NaBH{sub 4} hydrolysis reaction continued 200min in aluminum No 6061 alloy reactor with 0.3 wt% NaOH at 80-90 .deg. C.

Interactions of RuO4(g) with different surfaces in nuclear reactor containments

International Nuclear Information System (INIS)

Holm, J.; Glaenneskog, H.; Ekberg, C.

2008-07-01

During a severe nuclear reactor accident with air ingress, ruthenium in the form of RuO4 can be released from the nuclear fuel. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. This work has investigated the distribution of RuO4 between an aqueous and gaseous phase in the temperature interval of 20-50 deg. C by on-line measurements with an experimental set-up made of glass. The experiments showed that RuO4 is almost immediately distributed in the aqueous phase after its introduction in the set-up in the entire temperature interval. However, the deposition of ruthenium on the glass surfaces in the system was significant. The speciation of the ruthenium on the glass surfaces was studied by SEM-EDX and ESCA and was determined to be the expected RuO2. Experiments of interactions between gaseous ruthenium tetroxide and the metals aluminium, copper and zinc have been investigated. The metals were treated by RuO4 (g) at room temperature and analyzed with ESCA, SEM and XRD. The analyses show that the black ruthenium deposits on the metal surfaces were RuO2, i.e. the RuO4 (g) has been transformed on the metal surfaces to RuO2(s). The analyses showed also that there was a significant deposition of ruthenium tetroxide especially on the copper and zinc samples. Aluminium has a lower ability to deposit gaseous ruthenium tetroxide than the other metals. The conclusion that can be made from the results is that surfaces in nuclear reactor containments will likely reduce the source term in the case of a severe accident in a nuclear power plant. (au)

Evaluation of Tehran research reactor (TRR) control rod worth using MCNP4C computer code

International Nuclear Information System (INIS)

Hosseini, Mohammad; Vosoughi, Naser; Hosseini, Seyed Abolfazl

2010-01-01

The main objective of reactor control system is to provide a safe reactor starting up, operation and shutting down. Calculation or measurement of precise values of control rod worth is of great importance in Tehran Research Reactor (TRR), considering the fact that they are the only controlling tools in the reactor. In present paper, simulation of TRR in First Operation Cycle (FOC) and in cold and clean core for the calculation of total and integral worth of control nods is reported. MCNP4C computer code has been used for all simulation process. Two method have been used for control rods worth calculation in this paper, namely the direct approach and perturbation method. It is shown that while the direct approach is appropriate for worth calculation of both the shim and the regulating control rods, the perturbation method is just suitable for tiny reactivity changes, i.e. for small initial part of regulating rods. Results of simulation are compared with the reported data in Safety Analysis Report (SAR) of Tehran research reactor and showed satisfactory agreement. (author)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

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

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

Nizet Jean, Pichault François, Les performances des organisations africaines : pratiques de gestion en contexte incertain

Directory of Open Access Journals (Sweden)

David Laloy

2011-02-01

Full Text Available Cet ouvrage, dirigé par Jean Nizet et François Pichault, propose d’aborder la question du fonctionnement et des performances des organisations africaines en dépassant la posture “radicale” selon laquelle les caractéristiques de la culture africaine (manque d’expérience et de formation des entrepreneurs, clientélisme et corruption, prégnance des cultures traditionnelles et non rationnelles... expliqueraient à elles seules l’inhibition du développement des organisations. Ce type d’approche, qu...

Research reactor utilization. Summary reports of three study group meetings: Irradiation techniques at research reactors, held in Istanbul 15-19 November 1965; Research reactor operation and maintenance problems, held in Caracas 6-10 December 1965; and Research reactor utilization in the Far East, held in Lucas Heights 28 February - 4 March 1966

International Nuclear Information System (INIS)

1967-01-01

The three sections of this book, which are summary reports of three Study Group meetings of the IAEA: Irradiation techniques at research reactors, Istanbul, 15-19 November 1965; Research reactor operation and maintenance problems, Caracas, 6-10 December 1965; and Research reactor utilization in the Far East, Lucas Heights, Australia, 28 February - 4 March 1966. These meetings were the latest in a series designed to promote efficient utilization of research reactors, to disseminate information on advances in techniques, to discuss common problems in reactor operations, and to outline some advanced areas of reactor-based research. (author)

The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

2007-01-01

CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application in dispersed energy supply and multipurpose use, in conjunction with Toshiba Corporation. The 4S is sodium cooled fast reactor and their electrical output has two options of 10MWe and 50MWe. In this paper, 10MWe 4S (4S-10M) was proposed. 4S-10M has some unique features. It employs a burn-up control system with annular reflector in place of the control rod that requires the frequent maintenance service. The core life time of the 4S-10M is 30 years and the fuel transport is not required during core life time. All temperature feedback coefficients are negative during core life time. In the latest design for 4S-10M, a pool and tall type reactor design was selected to reduce the construction cost. Two types of decay heat removal system (Reactor Vessel Auxiliary Cooling System; RVACS, Intermediate Reactor Auxiliary Cooling System; IRACS) using natural convection power were adopted. It is necessary to confirm that these two heat removal system can operate appropriately. The transition analyses were executed by the CERES code to evaluate the design feasibility and the thermal hydraulic characteristics of the 4S-10M. CERES is a multi-dimensional plant dynamics simulation code for liquid metal reactors developed by the CRIEPI. CERES can perform simulations ranging from forced circulation (full/partial power operation) to natural circulation. Components (pumps, IHXs, SGs, pipings, etc.) of the reactor are modeled as one-dimensional. Multi-dimensional plena are connected to such components. Two loss-of-flow accident sequences are considered. In the first case, it is assumed that the primary and the secondary pump were stopped by the total station black out. The reactor shut down system was assumed to be success. This sequence is referred to as the protected loss-of-flow accident (PLOF). In the second case, it is assumed that the reactor shut down systems fail to operate and the

O inventário das curiosidades botânicas da Nouvelle France de Pierre-François-Xavier de Charlevoix (1744 The inventory of botanical curiosities in Pierre-François-Xavier de Charlevoix's Nouvelle France (1744

Directory of Open Access Journals (Sweden)

Michel Kobelinski

2013-03-01

Full Text Available Verifica a extensão dos aportes botânicos de Pierre-François-Xavier de Charlevoix em Histoire et description générale de la Nouvelle France em relação a trabalhos de pesquisadores anteriores, suas valorações das representações iconográficas e discursivas e aplicabilidade no projeto de colonização francesa. Investiga-se o que o levou a preterir o modelo taxionômico de Lineu e o que pretendia com seu catálogo de curiosidades botânicas. O desenlace de sua trajetória filosófico-religiosa permite compreender seu posicionamento no quadro de classificação da natureza, os sentidos das informações etnológicas, as formas de apropriação intelectual e os usos da iconografia botânica e do discurso como propaganda político-emotiva para incentivar a ocupação colonial.The article explores the botanical contributions of Pierre-François-Xavier de Charlevoix's book Histoire et description générale de la Nouvelle France vis-à-vis the contributions of previous researchers, his use of iconographic and discursive representations and its relevance to the project of French colonization. It investigates why he refused Linnaeus' taxonomic model and what he intended with his catalogue of botanical curiosities. The unfolding of his philosophical and religious trajectory allows to understand his stance regarding the classification of nature, the meanings of ethnological information, his forms of intellectual appropriation, and his use of discourse and botanical iconography as political and emotional propaganda to encourage colonial settlement.

Troubles detected during regular inspection of No.1 plant in Oi Power Station, Kansai Electric Power Co., Inc

International Nuclear Information System (INIS)

1990-01-01

No. 1 plant in Oi Power Station, Kansai Electric Power Co., Inc. is a PWR plant with rated output of 1175 MW, and its regular inspection is carried out since August 14, 1989. When eddy current flaw detection inspection was carried out on the total number (11426 except already plugged tubes) of the heating tubes of steam generators, significant indication was observed in tube supporting plate part of 279 tubes, at the boundary of tube plate expanded part of 34 tubes, and in the tube plate expanded part of 99 tubes, 411 heating tubes in total (all on high temperature side). Consequently, it was decided to repair 367 tubes using sleeves, and to plug other 44 tubes. Besides, among the heating tubes plugged in the past, it was decided to remove plugs from 161 tubes, and by repairing them with sleeves, to use them again. Total number of heating tubes 13552 (3388 tubes x 4 steam generators), Number of plugged tubes 2009 (decrease by 117 this time), Ratio of plugging 14.8%. (K.I.)

FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

International Nuclear Information System (INIS)

1996-09-01

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option

The EL-4 reactor. Changing of a pressure tube on a test loop

International Nuclear Information System (INIS)

Foulquier, H.; Clara, P.

1964-01-01

Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2007

International Nuclear Information System (INIS)

2012-03-01

In the fiscal year 2007, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) and the JRR-4 was operated for 92 days. JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 262 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analyses, prompt gamma-ray analyses, and others submitted by the users in JAEA and other Organizations. (author)

Canadian supercritical water reactor modeling using G4STORK

International Nuclear Information System (INIS)

Ford, W.; Buijs, A.

2015-01-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Canadian supercritical water reactor modeling using G4STORK

Energy Technology Data Exchange (ETDEWEB)

Ford, W.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

2015-07-01

The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2008-01-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2012-01-01

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Science.gov (United States)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

Energy Technology Data Exchange (ETDEWEB)

Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

2012-01-15

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

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

Operating reactors licensing actions summary. Volume 4, No. 9

International Nuclear Information System (INIS)

1984-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the division of licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Vol. 4, No. 2

International Nuclear Information System (INIS)

1984-04-01

This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

The Chernobyl-4 Reactor and the possible causes of the accident

International Nuclear Information System (INIS)

Motte, F.

1986-01-01

A description and information about the Chernobyl nuclear reactor is given. Some comparison elements between the RBMK reactor type and GCR, CANDU, SGHWR and Hanford N reactor types are presented. A scenario of the possible causes of the accident is discussed. (A.F.)

Organic-inorganic semiconductor devices and 3, 4, 9, 10 perylenetetracarboxylic dianhydride: an early history of organic electronics

International Nuclear Information System (INIS)

Forrest, S R

2003-01-01

The demonstration, over 20 years ago, of an organic-inorganic heterojunction (OI HJ) device along with investigations of the growth and physical properties of the archetypal crystalline molecular organic semiconductor 3, 4, 9, 10 perylenetetracarboxylic dianhydride are discussed. Possible applications of OI HJ devices are introduced and the dramatic change in conductive properties of these materials when exposed to high-energy ion beams is described. The past and future prospects for hybrid organic-on-inorganic semiconductor structures for use in electronic and photonic applications are also presented

Present status of decommissioning in the Musashi Reactor Facility (4)

International Nuclear Information System (INIS)

Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

2012-01-01

The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

Calculations of fuel burn up and radionuclide inventories in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION codes

International Nuclear Information System (INIS)

Khattab, K.

2005-01-01

The WIMSD4 code is used to generate the fuel group constants and the infinite multiplication factor as a function of the reactor operating time for 10, 20, and 30 k W operating power levels. The uranium burn up rate and burn up percentage, the amounts of the plutonium isotopes, the concentrations and radioactivities of the fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core are calculated using the WIMSD4 code as well. The CITATION code is used to calculate the changes in the effective multiplication factor of the reactor.(author)

Nucleogenic radioiodination of O-iodo hippuric acid (O-I H A) VIA molten acetic acid analogs (A A A). Vol. 3

Energy Technology Data Exchange (ETDEWEB)

El-Shaboury, G; El-Kolaly, M T; El-Watery, A; El-Mohty, A; Raieh, M [Radioisotope Production and Labelled Compounds Department, Hot Laboratories Center, Atomic Energy Authority, Cairo, (Egypt)

1996-03-01

A recent study for nucleogenic radioiodination of O-iodo hippuric acid (O-I H A) in dry-state (i.e. Molten state) with radioiodine in molten acetic acid analogs (AAA) has been investigated. The result investigated has revealed that the molten ammonium acetate (m.p. 114 degree C) fulfills the desired requirements for achieving high and pure radiochemical yield up to 95% within 5 min. at 120 degree C, when used as a molten medium for the no-carrier added isotope - exchange reaction between inactive O-I H A and Lyophilized ethanolic solution of sodium iodide ({sup 131} I{sup -}). On the other hand, the different critical parameters which affects the isotopic - exchange reaction in molten state previously described are discussed to evaluate the chemical principles of the reaction. Also the product obtained is completely free from impurities currently found in commercial radioiodinated - hippuran usually obtained by molten techniques such as glycyl - O - iodihippuric acid (g-OIHA) as well as O-iodobenzonic acid (O-IBA), which are investigated by TIC silica G-60 using the organic phase of the following solvent consists of benzene: acetic acid: water: n.butanol in the ratio of 5:5:2:1 as developing solvent. 2 figs., 2 tabs.

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

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

A simulation of a pebble bed reactor core by the MCNP-4C computer code

Directory of Open Access Journals (Sweden)

Bakhshayesh Moshkbar Khalil

2009-01-01

Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

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

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

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

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

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

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

Nuclear reactor technology progress report, vol. 4

International Nuclear Information System (INIS)

1981-01-01

The works of the Engineering Section, Fast Experimental Reactor Division, are roughly classified into the technologies concerning the reactor core, abnormality monitoring, the plant, purity control and operation planning. In this paper, the activities of the Engineering Section, the operational results of Joyo and the foreign informations on FBRs in this quarter are reported. The second regular inspection carried out successively from the previous quarter was completed, and the fourth cycle operation of Joyo at 75 MW was started. The measurement of CP around the primary system pipings and equipments, the preliminary test of a core flow meter for Monju, and the various characteristic tests were carried out during this period. 2 N reports, 1 SA report and 63 memos were drawn up in this quarter. The test plan to be carried out during the period of the fourth to sixth cycle operations in this last year using the MK-1 core was formed and decided. Various meetings within and outside the division are reported. The data obtained in the operational characteristic test and special test are shown. As the results concerning the reactor technologies, the development of dosimetry techniques, the measurement and analysis of the core characteristics, the measurement of the temperature and flow velocity of coolant at the fuel assembly exit, the system pressure loss in the primary cooling system and others are reported. (Kako, I.)

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

Energy Technology Data Exchange (ETDEWEB)

Guentay, S [ed.; Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs.

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

International Nuclear Information System (INIS)

Guentay, S.

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs

The accident of Chernobylsk-4 reactor and its consequences

International Nuclear Information System (INIS)

1986-01-01

This report deals with the particulars of the accident as communicated by the Soviet delegation at an IAEA meeting by the and of August 1986. It was stated that the consequences emanated from the inherent instability of the design of the reactor, the deviation from the safety rules by the operators and the lack of a sight reactor containment. (G.B.)

Les discours de François Hollande (2009-2012: la force axiologico-affective du changement

Directory of Open Access Journals (Sweden)

Maria Immacolata Spagna

2014-03-01

Full Text Available Abstract – Within the field of argumentative analysis, our purpose is to underline the function and the effectiveness of the axiological-emotional content of the change proposed by François Hollande in his speeches (2009-2012.On the basis of the emotional orientation, given in a dysphoric register towards the previous government and in a euphoric one towards the auspicious future, the argumentation of the change claimed by the “normal” president is based on values. To not change would mean to accept the current situation and therefore all its negative values.Putting the reader in a condition of emotional tension with the aim of energizing him to find a remedy, Hollande’s proposition becomes, thus, a call to action, an implicit request for social, politic and moral engagement, to change the course of history and to realize a better future. Keywords: Argumentation, political discourse, emotion, axiological, change.  Résumé – Dans le cadre de l’analyse argumentative, cet article vise à souligner la fonction et l’efficacité de la charge axiologico-affective du changement proposé par François Hollande dans ses discours (2009-2012.Sur la base de l’orientation émotionnelle, donnée dans un registre dysphorique à l’égard du gouvernement passé et euphorique vers l’avenir prometteur, l’argumentation du changement invoqué par le président “normal” se fonde sur les valeurs. Ne pas changer signifierait accepter la situation actuelle et, par là, toutes ces valeurs négatives.Mettant le lecteur dans une condition de tension émotive pour le motiver à trouver un remède, la proposition de Hollande devient ainsi un appel à agir, une requête implicite d’engagement social, politique et moral pour changer le cours de l’histoire et réaliser un futur meilleur. Mots clés: argumentation, discours politique, émotion, axiologique, changement. 

Magnetic susceptibility as a direct measure of oxidation state in LiFePO4 batteries and cyclic water gas shift reactors.

Science.gov (United States)

Kadyk, Thomas; Eikerling, Michael

2015-08-14

The possibility of correlating the magnetic susceptibility to the oxidation state of the porous active mass in a chemical or electrochemical reactor was analyzed. The magnetic permeability was calculated using a hierarchical model of the reactor. This model was applied to two practical examples: LiFePO4 batteries, in which the oxidation state corresponds with the state-of-charge, and cyclic water gas shift reactors, in which the oxidation state corresponds to the depletion of the catalyst. In LiFePO4 batteries phase separation of the lithiated and delithiated phases in the LiFePO4 particles in the positive electrode gives rise to a hysteresis effect, i.e. the magnetic permeability depends on the history of the electrode. During fast charge or discharge, non-uniform lithium distributionin the electrode decreases the hysteresis effect. However, the overall sensitivity of the magnetic response to the state-of-charge lies in the range of 0.03%, which makes practical measurement challenging. In cyclic water gas shift reactors, the sensitivity is 4 orders of magnitude higher and without phase separation, no hysteresis occurs. This shows that the method is suitable for such reactors, in which large changes of the magnetic permeability of the active material occurs.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2007. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Miyazaki, Osamu; Awa, Yasuaki; Isaka, Koji; Kutsukake, Kenichi; Komeda, Masao; Shibata, Ko; Hiyama, Kazuhisa; Suzuki, Mayu; Sone, Takuya; Ohuchi, Tomoaki; Terakado, Yuichi; Sataka, Masao

2009-06-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor-3), JRR-4(Japan Research Reactor-4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2007 and March 31, 2008. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator. (2) Utilization of research reactors and tandem accelerator. (3) Upgrading of utilization techniques of research reactors and tandem accelerator. (4) Safety administration for research reactors and tandem accelerator. (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, plans and outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2010. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and Tandem Accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Yamada, Yusuke; Kawashima, Kazuhiro; Asozu, Takuhiro; Nakamura, Takemi; Arai, Masaji; Yoshinari, Shuji; Sataka, Masao

2012-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2010 and March 31, 2011. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for research reactors and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, commendation, outcomes in service and technical developments and so on. (author)

Generalities about nuclear reactors

International Nuclear Information System (INIS)

Jaouen, C.; Beroux, P.

2012-01-01

From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2011. Operation, utilization and technical development of JRR-3, JRR-4, NSRR and tandem accelerator

International Nuclear Information System (INIS)

Ishii, Tetsuro; Nakamura, Kiyoshi; Kawamata, Satoshi; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Kabumoto, Hiroshi; Nakamura, Takemi; Tamura, Itaru; Kawasaki, Sayuri; Sataka, Masao

2013-03-01

The Department of Research Reactors and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor) and Tandem Accelerator. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2011 and March 31, 2012. The activities were categorized into six service/development fields: (1) Recovery from the Great East Japan Earthquake, (2) Operation and maintenance of research reactors and tandem accelerator, (3) Utilization of research reactors and tandem accelerator, (4) Upgrading of utilization techniques of research reactors and tandem accelerator, (5) Safety administration for research reactors and tandem accelerator, (6) International cooperation. Also contained are lists of publications, meetings, granted permissions on lows and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, commendation, outcomes in service and technical developments and so on. (author)

An initial assessment of the Chernobyl-4 reactor accident release source

International Nuclear Information System (INIS)

Macdonald, H.F.; ApSimon, H.M.; Wilson, J.J.N.

1986-07-01

The long-range atmospheric dispersion model MESOS has been used to provide a preliminary evaluation of the effects over Western Europe of radioactivity released during the accident which occurred at the Chernobyl-4 reactor in the USSR in April 1986. The results of this analysis have been compared with observations during the first week or so following the accident of airborne contamination levels at a range of locations across Europe in order to obtain an estimate of accident release source. The work presented here was performed during the 6-8 weeks following the accident and the results obtained will be subject to refinement as more detailed data become available. However, at this early stage they indicate a release source for the Chernobyl accident, expressed as a fraction of the estimated reactor core inventory, of approx. 15-20% of the iodine and caesium isotopes, approx. 1% of the ruthenium and lesser amounts of the other fission products and actinides, together with an implied major fraction of the krypton and xenon noble gases. (author)

Examination policy concerning the additional installation of No. 3 and No. 4 reactors in Takahama Nuclear Power Station and No. 3 and No. 4 reactors in Fukushima No. 2 Nuclear Power Station

International Nuclear Information System (INIS)

1980-01-01

The Nuclear Safety Commission decided the annual examination policy on the modification of reactor installation in Takahama Nuclear Power Station to construct No. 3 and No. 4 reactors inquired under date of November 26, 1979, by the Minister of International Trade and Industry, so that the examination results of the accident in Three Mile Island nuclear power station are reflected to the examination for the purpose of improving reactor safety. The examination results of the accident in Three Mile Island power station are being investigated by the Committee on Examination of Reactor Safety, based on the policy shown in ''On the second report of the special committee examining the accident in a nuclear power station in the U.S.'' determined by the Nuclear Safety Commission under date of September 13, 1979. Though the Committee will further clarify the past guideline about the items concerning the criteria, design and operation management, the Committee decided the tentative policy to reflect it to safety examination. Further, a table is attached, in which 52 items to be reflected to the security measures are classified from the viewpoint of necessity to reflect them to the final examination. This table includes 13 items of criteria and examination, 7 items related to design, 10 items related to operation management, 10 antidisaster items, and 12 items related to safety research. (Wakatsuki, Y.)

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

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

Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

International Nuclear Information System (INIS)

Khattab, K.; Omar, H.; Ghazi, N.

2009-01-01

A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

Assimilation of the AVISO Altimetry Data into the Ocean Dynamics Model with a High Spatial Resolution Using Ensemble Optimal Interpolation (EnOI)

Science.gov (United States)

Kaurkin, M. N.; Ibrayev, R. A.; Belyaev, K. P.

2018-01-01

A parallel realization of the Ensemble Optimal Interpolation (EnOI) data assimilation (DA) method in conjunction with the eddy-resolving global circulation model is implemented. The results of DA experiments in the North Atlantic with the assimilation of the Archiving, Validation and Interpretation of Satellite Oceanographic (AVISO) data from the Jason-1 satellite are analyzed. The results of simulation are compared with the independent temperature and salinity data from the ARGO drifters.

Homogeneity of Continuum Model of an Unsteady State Fixed Bed Reactor for Lean CH4 Oxidation

Directory of Open Access Journals (Sweden)

Subagjo

2014-07-01

Full Text Available In this study, the homogeneity of the continuum model of a fixed bed reactor operated in steady state and unsteady state systems for lean CH4 oxidation is investigated. The steady-state fixed bed reactor system was operated under once-through direction, while the unsteady-state fixed bed reactor system was operated under flow reversal. The governing equations consisting of mass and energy balances were solved using the FlexPDE software package, version 6. The model selection is indispensable for an effective calculation since the simulation of a reverse flow reactor is time-consuming. The homogeneous and heterogeneous models for steady state operation gave similar conversions and temperature profiles, with a deviation of 0.12 to 0.14%. For reverse flow operation, the deviations of the continuum models of thepseudo-homogeneous and heterogeneous models were in the range of 25-65%. It is suggested that pseudo-homogeneous models can be applied to steady state systems, whereas heterogeneous models have to be applied to unsteady state systems.

Evaluation of the OSCAR-4/MCNP calculation methodology for radioisotope production in the SAFARI-1 reactor

International Nuclear Information System (INIS)

Karriem, Z.; Zamonsky, O.M.

2014-01-01

The South African Nuclear Energy Corporation SOC Ltd (Necsa) is a state owned nuclear facility which owns and operates SAFARI-1, a 20 MW material testing reactor. SAFARI-1 is a multi-purpose reactor and is used for the production of radioisotopes through in-core sample irradiation. The Radiation and Reactor Theory (RRT) Section of Necsa supports SAFARI-1 operations with nuclear engineering analyses which include core-reload design, core-follow and radiation transport analyses. The primary computer codes that are used for the analyses are the OSCAR-4 nodal diffusion core simulator and the Monte Carlo transport code MCNP. RRT has developed a calculation methodology based on OSCAR-4 and MCNP to simulate the diverse in-core irradiation conditions in SAFARI-1, for the purpose of radioisotope production. In this paper we present the OSCAR-4/MCNP calculation methodology and the software tools that were developed for rapid and reliable construction of MCNP analysis models. The paper will present the application and accuracy of the methodology for the production of yttrium-90 ( 90 Y) and will include comparisons between calculation results and experimental measurements. The paper will also present sensitivity analyses that were performed to determine the effects of control rod bank position, representation of core depletion state and sample loading configuration, on the calculated 90 Y sample activity. (author)

The Cartesian doctor, François Bayle (1622-1709), on psychosomatic explanation.

Science.gov (United States)

Easton, Patricia

2011-06-01

There are two standing, incompatible accounts of Descartes' contributions to the study of psychosomatic phenomena that pervade histories of medicine, psychology, and psychiatry. The first views Descartes as the father of "rational psychology" a tradition that defines the soul as a thinking, unextended substance. The second account views Descartes as the father of materialism and the machine metaphor. The consensus is that Descartes' studies of optics and motor reflexes and his conception of the body-machine metaphor made early and important contributions to physiology and neuroscience but otherwise his impact was minimal. These predominately negative assessments of Descartes' contributions give a false impression of the role his philosophy played in the development of medicine and psychiatry in seventeenth-century France and beyond. I explore Descartes' influence in the little-known writings of a doctor from Toulouse, François Bayle (1622-1709). A study of Bayle gives us occasion to rethink the nature and role of psychosomatic explanation in Descartes' philosophy. The portrait I present is of a Cartesian science that had an actual and lasting effect on medical science and practice, and may offer something of value to practitioners today. Copyright © 2010 Elsevier Ltd. All rights reserved.

Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

Energy Technology Data Exchange (ETDEWEB)

El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

International Nuclear Information System (INIS)

El-Messeiry, A.M.

1996-01-01

In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

Calculation of the effective delayed neutron fraction by TRIPOLI-4 code for IPEN/MB-01 research reactor

International Nuclear Information System (INIS)

Lee, Y.K.; Hugot, F.X.

2011-01-01

The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)

Generation IV reactors: economics

International Nuclear Information System (INIS)

Dupraz, B.; Bertel, E.

2003-01-01

The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

François Arago a 19th century French humanist and pioneer in astrophysics

CERN Document Server

Lequeux, James

2016-01-01

François Arago, the first to show in 1810 that the surface of the Sun and stars is made of incandescent gas and not solid or liquid, was a prominent physicist of the 19th century. He used his considerable influence to help Fresnel, Ampere and others develop their ideas and make themselves known. This book covers his personal contributions to physics, astronomy, geodesy and oceanography, which are far from negligible, but insufficiently known. Arago was also an important and influential political man who, for example, abolished slavery in the French colonies. One of the last humanists, he had a very broad culture and range of interests. In parallel to his biography, this title also covers the spectacular progresses of science at the time of Arago, especially in France: the birth of physical optics, electromagnetism and thermodynamics. Francois Arago’s life is a fascinating epic tale that reads as a novel.

Molten salt reactor concept

International Nuclear Information System (INIS)

Sood, D.D.

1980-01-01

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

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2011-01-01

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR.

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system

Energy Technology Data Exchange (ETDEWEB)

Kuscu, Ozlem Selcuk, E-mail: oselcuk@mmf.sdu.edu.tr [Department of Environmental Engineering, Engineering and Architecture Faculty, Sueleyman Demirel University, Cuenuer Campus, 32260 Isparta (Turkey); Sponza, Delia Teresa [Dokuz Eyluel University, Engineering Faculty, Environmental Engineering Department, Buca Kaynaklar campus, Izmir (Turkey)

2011-03-15

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR.

Application of Box-Wilson experimental design method for 2,4-dinitrotoluene treatment in a sequential anaerobic migrating blanket reactor (AMBR)/aerobic completely stirred tank reactor (CSTR) system.

Science.gov (United States)

Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

2011-03-15

A sequential aerobic completely stirred tank reactor (CSTR) following the anaerobic migrating blanket reactor (AMBR) was used to treat a synthetic wastewater containing 2,4-dinitrotoluene (2,4-DNT). A Box-Wilson statistical experiment design was used to determine the effects of 2,4-DNT and the hydraulic retention times (HRTs) on 2,4-DNT and COD removal efficiencies in the AMBR reactor. The 2,4-DNT concentrations in the feed (0-280 mg/L) and the HRT (0.5-10 days) were considered as the independent variables while the 2,4-DNT and chemical oxygen demand (COD) removal efficiencies, total and methane gas productions, methane gas percentage, pH, total volatile fatty acid (TVFA) and total volatile fatty acid/bicarbonate alkalinity (TVFA/Bic.Alk.) ratio were considered as the objective functions in the Box-Wilson statistical experiment design in the AMBR. The predicted data for the parameters given above were determined from the response functions by regression analysis of the experimental data and exhibited excellent agreement with the experimental results. The optimum HRT which gave the maximum COD (97.00%) and 2,4-DNT removal (99.90%) efficiencies was between 5 and 10 days at influent 2,4-DNT concentrations 1-280 mg/L in the AMBR. The aerobic CSTR was used for removals of residual COD remaining from the AMBR, and for metabolites of 2,4-DNT. The maximum COD removal efficiency was 99% at an HRT of 1.89 days at a 2,4-DNT concentration of 239 mg/L in the aerobic CSTR. It was found that 280 mg/L 2,4-DNT transformed to 2,4-diaminotoluene (2,4-DAT) via 2-amino-4-nitrotoluene (2-A-4-NT) and 4-amino-2-nitrotoluene (4-A-2-NT) in the AMBR. The maximum 2,4-DAT removal was 82% at an HRT of 8.61 days in the aerobic CSTR. The maximum total COD and 2,4-DNT removal efficiencies were 99.00% and 99.99%, respectively, at an influent 2,4-DNT concentration of 239 mg/L and at 1.89 days of HRT in the sequential AMBR/CSTR. Copyright © 2011 Elsevier B.V. All rights reserved.

Technology assessment HTR. Part 4. Power upscaling of High Temperature Reactors

International Nuclear Information System (INIS)

Van Heek, A.I.

1996-06-01

Designs of nuclear reactors can be classified in evolutionary, revolutionary and innovative designs. An innovative design is the High Temperature Reactor (HTR). Introduction of innovative reactors has not been successful until now. Globally, three requirements for this reactors for successful market introduction can be identified: (1) Societal support for nuclear energy, or if separable, for this reactor type, should be repaired; (2) After market introduction the innovative plant must be able to operate economically competitive; and (3) The costs of market introduction of an innovative reactor design must be limited. Until now all reactor designs classified as innovative have not yet been realized. High temperature reactors exist in many different designs. Common features are: helium coolant, graphite moderator and coated particle fuel. The combination of these creates the potential to fulfill the first requirement (public support), and similarly a hurdle to the second requirement (economical operation). All three problems existing in the eyes of the public are addressed, while a high degree of transparency is reached, making the design understandable also by others than nuclear experts. A consequence of designing according to the social support requirement is a limitation of the unit power level. The usual method to make nuclear power plants economically competitive, i.e. just raising the power level (economy of scale) could not be applied anymore. Therefore other means of cost decreasing had to be used: modularization and simplification. These ideas are explained. Since all existing HTRs are currently out of operation, additional experience from two small HTRs under construction at this moment in the Far East will be essential. In the history of HTR designs, an evolutionary path can be identified. The early designs had a philosophy of safety and economics very similar to those of LWR. Modularization was introduced to attain economic viability and the design was

A founder mutation in LEPRE1 carried by 1.5% of West Africans and 0.4% of African Americans causes lethal recessive osteogenesis imperfecta.

Science.gov (United States)

Cabral, Wayne A; Barnes, Aileen M; Adeyemo, Adebowale; Cushing, Kelly; Chitayat, David; Porter, Forbes D; Panny, Susan R; Gulamali-Majid, Fizza; Tishkoff, Sarah A; Rebbeck, Timothy R; Gueye, Serigne M; Bailey-Wilson, Joan E; Brody, Lawrence C; Rotimi, Charles N; Marini, Joan C

2012-05-01

Deficiency of prolyl 3-hydroxylase 1, encoded by LEPRE1, causes recessive osteogenesis imperfecta (OI). We previously identified a LEPRE1 mutation exclusively in African Americans and contemporary West Africans. We hypothesized that this allele originated in West Africa and was introduced to the Americas with the Atlantic slave trade. We aimed to determine the frequency of carriers for this mutation among African Americans and West Africans, and the mutation origin and age. Genomic DNA was screened for the mutation using PCR and restriction digestion, and a custom TaqMan genomic single-nucleotide polymorphism assay. The mutation age was estimated using microsatellites and short tandem repeats spanning 4.2 Mb surrounding LEPRE1 in probands and carriers. Approximately 0.4% (95% confidence interval: 0.22-0.68%) of Mid-Atlantic African Americans carry this mutation, estimating recessive OI in 1/260,000 births in this population. In Nigeria and Ghana, 1.48% (95% confidence interval: 0.95-2.30%) of unrelated individuals are heterozygous carriers, predicting that 1/18,260 births will be affected with recessive OI, equal to the incidence of de novo dominant OI. The mutation was not detected in Africans from surrounding countries. All carriers shared a haplotype of 63-770 Kb, consistent with a single founder for this mutation. Using linkage disequilibrium analysis, the mutation was estimated to have originated between 650 and 900 years before present (1100-1350 CE). We identified a West African founder mutation for recessive OI in LEPRE1. Nearly 1.5% of Ghanians and Nigerians are carriers. The estimated age of this allele is consistent with introduction to North America via the Atlantic slave trade (1501-1867 CE).

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

1999-04-01

The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

Gravity wave generation and propagation during geomagnetic storms over Kiruna (67.8°N, 20.4°E

Directory of Open Access Journals (Sweden)

P. R. Fagundes

1995-04-01

Full Text Available Atmospheric gravity waves, detected over Kiruna (67.8°N, 20.4°E during geomagnetic storms, are presented and analysed. The data include direct measurements of the OI 630.0 nm emission line intensity, the x-component of the local geomagnetic field and thermospheric (meridional and zonal wind velocities derived from the OI 630.0 nm Doppler shift observed with an imaging Fabry-Perot interferometer (IFPI. A low pass band filter technique was used to determine short-period variations in the thermospheric meridional wind velocities observed during geomagnetic storms. These short-period variations in the meridional wind velocities, which are identified as due to gravity waves, are compared to the corresponding variations observed in the OI 630.0 nm emission line intensity, x-component of the local geomagnetic field and the location of the auroral electrojet. A cross-correlation analysis was used to calculate the propagation velocities of the observed gravity waves.

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

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

Problems of creating fuel elements for fast gas-cooled reactors working on N2O4-dissociating coolant

International Nuclear Information System (INIS)

Nesterenko, V.B.; Zelensky, V.F.; Kolykhan, L.I.; Karpenko, G.V.; Krasnorutsky, V.S.; Isakov, V.P.; Ashikhmin, V.P.; Permyakov, L.N.

1985-01-01

A variant of fast gas-cooled reactors is one using dissociating N 2 O 4 nitrogen tetroxide as a coolant. This type of reactors is promising because of great thermal effects of dissociation reactions while heating and recombination while cooling; small latent heat of evaporation; high heat transfer coefficient owing to additional heat transfer in a chemical reaction; high N 2 O 4 density in a gas state at operation parameters. The mentioned advantages give possibility to create a small turbine, heat exchange apparatus and to get high heat production in the active zone. All this opens new ways to increase power plants effectiveness

Medical aspects of the nuclear accident in the Chernobylsk-4 reactor

International Nuclear Information System (INIS)

Arndt, D.; Schmidt, W.

1989-01-01

The Kiev conference on the Chernobylsk reactor accident was concerned with the following items: (1) Medical consequences and organization of medical assistance as well as aftercare of radiation-exposed persons. (2) Analysis of the postirradiation situation and judgement of the consequences of the accident as to the USSR population. (3) Peculiarities of external and internal radiation exposure of the population in the area controlled. (4) Organization and efficiency of the epidemiological register of the USSR. (5) Organization and judgement of educational work and public relations concerning the sanitary conditions in populations exposed to an increased contamination

Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca

International Nuclear Information System (INIS)

Novakovic, M.

1961-12-01

The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO 2 with beryllium cladding, cooled by CO 2 under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO 2 . This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment

Effect of Utilization of Silicide Fuel with the Density 4.8 gU/cc on the Kinetic Parameters of RSG-GAS Reactor

International Nuclear Information System (INIS)

Setiyanto; Sembiring, Tagor M.; Pinem, Surian

2007-01-01

Presently, the RSG-GAS reactor using silicide fuel element of 2.96 gU/cc. For increasing reactor operation time, its planning to change to higher density fuel. The kinetic calculation of silicide core with density 4.8 gU/cc has been carried out, since it has an influence on the reactor operation safety. The calculated kinetic parameters are the effective delayed neutron fraction, the delayed neutron decay constant, prompt neutron lifetime and feedback reactivity coefficient very important for reactor operation safety. the calculation is performed in 2-dimensional neutron diffusion-perturbation method using modified Batan-2DIFF code. The calculation showed that the effective delayed neutron fraction is 7. 03256x10 -03 , total delay neutron time constant is 7.85820x10 -02 s -1 and the prompt neutron lifetime is 55.4900 μs. The result of prompt neutron lifetime smaller 10 % compare with silicide fuel of 4.8 gU/cc. The calculated results showed that all of the feedback reactivity coefficient silicide core 4.8 gU/cc is negative. Totally, the feedback reactivity coefficient of silicide fuel of 4.8 gU/cc is 10% less than that of silicide fuel of 2.96 gU/cc. The results shown that kinetic parameters result decrease compared with the silicide core with density 2.96 gU/cc, but no significant influence in the RSG-GAS reactor operation. (author)

Elements on reactor control

International Nuclear Information System (INIS)

Bruna, G.B.

1998-01-01

In order to achieve the two-fold goal of maximizing the energy obtained from reactor fuel and ensuring the large flexibility of plant operation in respect to safety regulations and keeping the reactor integrity the control of PWRs is generally based on real time monitoring and analysing of independent neutronic parameters: thermal power release, axial power distribution in the core and temperatures of the primary loop. Two control chains more or less coupled according to the control chosen mode are in charge of the control of these parameters. With the brief history of control in French power reactors the advanced X control mode adopted by Framatome for N4 plants is described in detail. A summary of N4 reactor control and protection system is included

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

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

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

Supercritical Water Reactors

International Nuclear Information System (INIS)

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

2014-01-01

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

4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

International Nuclear Information System (INIS)

2012-12-01

The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

The development of octagon Zr-4 alloy tube for heating reactors

International Nuclear Information System (INIS)

Yang Fanglin; Yang Yingli; Wang Guangshen

1989-10-01

The asymmetrical octagon Zr-4 alloy tubes which are used for fuel assembly in the heating reactor have been developed. The thickness of tube wall is 1.5 mm and the length is 1725 mm. The long side of the octagon is 138.7 0.3 +0.2 mm, the short side is 93.1 ± 0.1 mm. To manufacture these tubes a stretch draw forming processing method is adopted. The process is divided into two phases. In the first phase, a short draw mould is used to stretch the Zr-4 alloy tube. In the second phase, a long draw mould, its length is equal to the end-produt length, is used to complete the final processing. The size accuracy and repeatability of this method are excellent and can fully meet the design requirements

Annex VII - Diagrams: 1. Reactor operation (1960-1977); 2. Mean daily reactor power density in 1977; 3. Monthly reactor power for 1977; 4. percent of utilization of experimental space in 1977; Prilog VII - Dijagrami: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-12-15

This Annex includes the following diagrams: 1. Annual Reactor RA power production (MWh) for the period from 1960-1977; 2. Mean daily reactor power density MW in 1977; 3. Monthly reactor power production (MWh) for 1977; 4. percent of utilization of experimental space in 1977. [Serbo-Croat] Ovaj prilog sadrzi dijagrame: 1. Rad reaktora (MWh) po godinama (1960-1977); 2. Srednja dnevna snaga reaktora u 1977. godini; 3. Rad reaktora (MWh) po mesecima za 1977. godinu i 4. Procenat iskoriscenja eksperimentalnog prostora u 1977. godini.

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2012. Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Murayama, Yoji; Ishii, Tetsuro; Nakamura, Kiyoshi; Uno, Yuki; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Odauchi, Shouji; Maruo, Takeshi

2014-03-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2012 and March 31, 2013. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for department of research reactor and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, outcomes in service and technical developments and so on. (author)

The molten salt reactor adventure

International Nuclear Information System (INIS)

MacPherson, H.G.

1985-01-01

A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF 4 -ThF 4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized

Hot electron bolometer heterodyne receiver with a 4.7-THz quantum cascade laser as a local oscillator

NARCIS (Netherlands)

Kloosterman, J.L.; Hayton, D.J.; Ren, Y.; Kao, T.Y.; Hovenier, J.N.; Gao, J.R.; Klapwijk, T.M.; Hu, Q.; Walker, C.K.; Reno, J.L.

2013-01-01

We report on a heterodyne receiver designed to observe the astrophysically important neutral atomic oxygen [OI] line at 4.7448?THz. The local oscillator is a third-order distributed feedback quantum cascade laser operating in continuous wave mode at 4.741?THz. A quasi-optical, superconducting NbN

Generation III+ Reactor Portfolio

International Nuclear Information System (INIS)

2010-03-01

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

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-08-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

First Calorimetric Measurement of OI-line in the Electron Capture Spectrum of $^{163}$Ho

CERN Document Server

Ranitzsch, P. C. -O.; Wegner, M.; Kempf, S.; Fleischmann, A.; Enss, C.; Gastaldo, L.; Herlert, A.; Johnston, K.

2014-01-01

The isotope $^{163}$Ho undergoes an electron capture process with a recommended value for the energy available to the decay, $Q_{\\rm EC}$, of about 2.5 keV. According to the present knowledge, this is the lowest $Q_{\\rm EC}$ value for electron capture processes. Because of that, $^{163}$Ho is the best candidate to perform experiments to investigate the value of the electron neutrino mass based on the analysis of the calorimetrically measured spectrum. We present for the first time the calorimetric measurement of the atomic de-excitation of the $^{163}$Dy daughter atom upon the capture of an electron from the 5s shell in $^{163}$Ho, OI-line. The measured peak energy is 48 eV. This measurement was performed using low temperature metallic magnetic calorimeters with the $^{163}$Ho ion implanted in the absorber. We demonstrate that the calorimetric spectrum of $^{163}$Ho can be measured with high precision and that the parameters describing the spectrum can be learned from the analysis of the data. Finally, we dis...

Characterizations of the diurnal shapes of OI 630.0 nm dayglow intensity variations: inferences

Directory of Open Access Journals (Sweden)

D. Chakrabarty

2002-11-01

Full Text Available Measurements of OI 630.0 nm thermospheric dayglow emission by means of the Dayglow Photometer (DGP at Mt. Abu (24.6° N, 73.7° E, dip lat 19.09° N, a station under the crest of Equatorial Ionization Anomaly (EIA, reveal day-to-day changes in the shapes of the diurnal profiles of dayglow intensity variations. These shapes have been characterized using the magnetometer data from equatorial and low-latitude stations. Substantial changes have been noticed in the shapes of the dayglow intensity variations between 10:00–15:00 IST (Indian Standard Time during the days when normal and counter electrojet events are present over the equator. It is found that the width (the time span corresponding to 0.8 times the maximum dayglow intensity of the diurnal profile has a linear relationship with the integrated electrojet strength. Occasional deviation from this linear relationship is attributed to the presence of substantial mean meridional wind.Key words. Ionosphere (equatorial ionosphere; ionosphere – atmosphere interactions; ionospheric disturbances

Characterizations of the diurnal shapes of OI 630.0 nm dayglow intensity variations: inferences

Directory of Open Access Journals (Sweden)

D. Chakrabarty

Full Text Available Measurements of OI 630.0 nm thermospheric dayglow emission by means of the Dayglow Photometer (DGP at Mt. Abu (24.6° N, 73.7° E, dip lat 19.09° N, a station under the crest of Equatorial Ionization Anomaly (EIA, reveal day-to-day changes in the shapes of the diurnal profiles of dayglow intensity variations. These shapes have been characterized using the magnetometer data from equatorial and low-latitude stations. Substantial changes have been noticed in the shapes of the dayglow intensity variations between 10:00–15:00 IST (Indian Standard Time during the days when normal and counter electrojet events are present over the equator. It is found that the width (the time span corresponding to 0.8 times the maximum dayglow intensity of the diurnal profile has a linear relationship with the integrated electrojet strength. Occasional deviation from this linear relationship is attributed to the presence of substantial mean meridional wind.

Key words. Ionosphere (equatorial ionosphere; ionosphere – atmosphere interactions; ionospheric disturbances

Management of spent fuel from research reactors - Brazilian progress report (within the framework of Regional Project IAEA-RLA-4/018)

International Nuclear Information System (INIS)

Soares, A.J.; Silva, J.E.R.

2005-01-01

There are four research reactors in Brazil. For three of them, because of the low reactor power and low burn-up of the fuel, except for the concern about ageing, spent fuel storage is not a problem. However for one of the reactors, more specifically IEA-R1 research reactor, the storage of spent fuel is a major concern, because, according to the proposed operation schedule for the reactor, unless an action is taken, by the year 2009 there will be no more racks available to store its spent fuel. This paper gives a brief description of the type and amount of fuel elements utilized in each one of the Brazilian research reactors, with a short discussion about the storage capacity at each installation. It also gives a description of the activities developed by Brazilian engineers and researchers during the period between 2001 and 2004, within the framework of regional project 'RLA-4/018-Management of Spent Fuel from Research Reactors'. As a conclusion, we can say that the advances of the project, and the integration promoted among the engineers and researchers of the participant countries were of fundamental importance for Brazilian researchers and engineers to understand the problems related to the storage of spent fuel, and to make a clear definition about the most suitable alternatives for interim storage of the spent fuel from IEAR1 research reactor. (author)

What occurred in the reactors

International Nuclear Information System (INIS)

Kudo, Kazuhiko

2013-01-01

Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

Compact torsatron reactors

International Nuclear Information System (INIS)

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

1988-05-01

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

A instrução pública na França revolucionária: considerações a partir do Essais sur l’enseignement en general et sur celui des mathématiques en particulier, de Sylvestre-François Lacroix - The public instruction in revolutionary France: some remarks accord

Directory of Open Access Journals (Sweden)

Antonio Vicente Marafioti Garnica

2013-04-01

Full Text Available O artigo aborda o livro Essais sur l'enseignement en général, et sur celui des mathématiques en particulier, de Sylvestre-François Lacroix, publicado originalmente na França em 1805, enfatizando sua primeira parte, na qual o autor discute a instrução pública e, particularmente, as escolas centrais da França revolucionária.Palavras-chave: Sylvestre-François Lacroix (1765-1843, instrução pública, Essais sur l'enseigne-ment en général, et sur celui des mathématiques en particulier, iluminismo, França. The public instruction in revolutionary France: some remarks according to Lacroix’s Essais sur l’enseignement en general et sur celui des mathématiques en particulierAbstractThe article discusses the book Essais sur l'enseignement en général, et sur celui des mathématiques en particulier, by Sylvestre-François Lacroix, originally published in France in 1805. The first part of the book, in which the author discusses public education, and particularly the central schools of revolutionary France, is emphasized.Key-words: Sylvestre-François Lacroix (1765-1843, public instruction, Essais sur l'enseignement en général, et sur celui des mathématiques en particulier, enlightenment, France. Instrucción pública en Francia revolucionaria: consideraciones a la vista del libro Essais sur l’enseignement en general et sur celui des mathématiques en particulier, de Sylvestre-François LacroixResumenEl artículo aborda el libro Essais sur l'enseignement en général, et sur celui des mathématiques en particulier de Sylvestre-François Lacroix, publicado originalmente en Francia, en 1805, haciendo hincapié en su primera parte, en la que el autor habla de la instrucción pública y en particular de las escuelas centrales en Francia en el período revolucionario.Palabras-clave: Sylvestre-François Lacroix (1765-1843, instrucción pública, Essais sur l'enseigne-ment en général, et sur celui des mathématiques en particulier

Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

International Nuclear Information System (INIS)

Bowman, S.M.; Suto, T.

1996-10-01

ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k eff of 1. 0040±0.0005

Fast reactors

International Nuclear Information System (INIS)

Vasile, A.

2001-01-01

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

Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

International Nuclear Information System (INIS)

Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

2017-01-01

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

Experiência e Utopia em Theodor W. Adorno, André Gorz e François Dubet

Directory of Open Access Journals (Sweden)

Sílvio César Camargo

2007-01-01

sociedade contemporânea. Conceito bastante complexo quanto a sua possibilidade de apreensão pela sociologia, experiência se refere tanto a problemas relativos ao conhecimento e suas possibilidades, mas também para as possibilidades de transformação da sociedade. Entende-se que há três teóricos da sociedade contemporânea, que partindo de bases epistemológicas diferentes, problematizam a emancipação humana a partir de uma singular atenção ao conceito de experiência. Theodor W. Adorno, André Gorz e François Dubet representam três formas de pensamento sobre a sociedade bastante diferentes, mas que possuem em comum o interesse normativo, e seu confronto nos mostra a importância do conceito de experiência para a compreensão da sociedade contemporânea e para pensar-se a utopia.

RA-0 reactor. New neutronic calculations; Reactor RA-0. Nuevos calculos neutronicos

Energy Technology Data Exchange (ETDEWEB)

Rumis, D; Leszczynski, F

1991-12-31

An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core`s interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author). [Espanol] En este trabajo se actualizan los calculos neutronicos realizados para el reactor RA-0, instalado en la Facultad de Ciencias Exactas, Fisicas y Naturales de la Universidad Nacional de Cordoba. Se describen los calculos realizados hasta la fecha y los resultados obtenidos. Las tecnicas incorporadas al calculo de un reactor como el RA-0 permiten predecir en detalle el comportamiento del flujo en el interior del nucleo y en el reflector, lo que sera una importante ayuda en el diseno de experimentos. En particular, el empleo del codigo WIMSD4 para calculos del reactor completo constituye una novedad en las posibles aplicaciones de ese codigo para resolver problemas que se presentan en la practica. (Autor).

Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

Examining the Relationship between Economic Hardship and Child Maltreatment Using Data from the Ontario Incidence Study of Reported Child Abuse and Neglect-2013 (OIS-2013

Directory of Open Access Journals (Sweden)

Rachael Lefebvre

2017-02-01

Full Text Available There is strong evidence that poverty and economic disadvantage are associated with child maltreatment; however, research in this area is underdeveloped in Canada. The purpose of this paper is to examine the relationship between economic hardship and maltreatment for families and children identified to the Ontario child protection system for a maltreatment concern. Secondary analyses of the Ontario Incidence Study of Reported Child Abuse and Neglect-2013 (OIS-2013 were conducted. The OIS-2013 examines the incidence of reported maltreatment and the characteristics of children and families investigated by child welfare authorities in Ontario in 2013. Descriptive and bivariate chi-square analyses were conducted in addition to a logistic regression predicting the substantiation of maltreatment. In 9% of investigations, the household had run out of money for food, housing, and/or utilities in the past 6 months. Children in these households were more likely to have developmental concerns, academic difficulties, and caregivers with mental health concerns and substance use issues. Controlling for key clinical and case characteristics, children living in families facing economic hardship were almost 2 times more likely to be involved in a substantiated maltreatment investigation (OR = 1.91, p < 0.001. The implications in regard to future research and promoting resilience are discussed.

The IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors

Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2001-01-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2001-04-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

International Nuclear Information System (INIS)

Johnson, A.B. Jr.

1976-09-01

Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

Validation gets underway on Sizewell ''Incredibility of Failure'' components

International Nuclear Information System (INIS)

Anon.

1992-01-01

The Inspection Validation Centre (IVC) of AEA Reactor Services in the UK has begun an eighteen month programme to validate the procedures and personnel of OIS plc, the inspection agents chosen by Nuclear Electric to carry out the pre-service ultrasonic inspection of the Sizewell B Pressurized Water Reactor components assigned to the ''Incredibility of Failure'' (IoF) category. The work involves several Sizewell B primary circuit components - the steam generators, pressurizer, and primary pumps - and will consider the inspections to be applied to the circumferential and nozzle-to-shell welds, nozzle inner radii and the pump fly-wheel forging. The validation will provide independent confirmation that OIS personnel are capable of using manual and automated methods to find and size any flaws of structural concern in these components. (author)

Studies of conceptual spheromak fusion reactors

International Nuclear Information System (INIS)

Katsurai, M.; Yamada, M.

1982-01-01

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A >or approx. 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type operation, (2) the ignited DT reactor, and (3) the ignited catalysed-type DD reactor. With a total wall loading of approx. 4 MW.m -2 , it is found that edge magnetic fields of only approx. 4 T (DT) and approx. 9 T (Cat. DD) are required for ignited reactors of 1 m plasma (minor) radius with output powers in the gigawatt range. An assessment of various schemes of generation, compression and translation of spheromak plasmas is presented. (author)

The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

Energy Technology Data Exchange (ETDEWEB)

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

2001-02-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

Hythane (H2 and CH4) production from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals via up-flow anaerobic self-separation gases reactor

International Nuclear Information System (INIS)

Mahmoud, Mohamed; Elreedy, Ahmed; Pascal, Peu; Sophie, Le Roux; Tawfik, Ahmed

2017-01-01

Highlights: • Bio-hythane production from polyester wastewater via UASG reactor was assessed. • Impacts of influent contamination by 1,4-dioxane and heavy metals were discussed. • Maximum volumetric H 2 and CH 4 productions of 0.12 and 1.06 L/L/d were achieved. • Significant drop in CH 4 production was resulted at OLR up to 1.07 ± 0.06 gCOD/L/d. • Bioenergy recovery through UASG economically achieved a net profit of 10,231 $/y. - Abstract: A long-term evaluation of hythane generation from unsaturated polyester resin wastewater contaminated by 1,4-dioxane and heavy metals was investigated in a continuous up-flow anaerobic self- separation gases (UASG) reactor inoculated with mixed culture. The reactor was operated at constant hydraulic retention time (HRT) of 96 h and different organic loading rates (OLRs) of 0.31 ± 0.04, 0.71 ± 0.08 and 1.07 ± 0.06 gCOD/L/d. Available data showed that volumetric hythane production rate was substantially increased from 0.093 ± 0.021 to 0.245 ± 0.016 L/L/d at increasing OLR from 0.31 ± 0.04 to 0.71 ± 0.08 gCOD/L/d. However, at OLR exceeding 1.07 ± 0.06 gCOD/L/d, it was dropped to 0.114 ± 0.016 L/L/d. The reactor achieved 1,4-dioxane removal efficiencies of 51.8 ± 2.8, 35.9 ± 1.6 and 26.3 ± 1.6% at initial 1,4-dioxane concentrations of 1.14 ± 0.28, 1.97 ± 0.41 and 4.21 ± 0.30 mg/L, respectively. Moreover, the effect and potential removal of the contaminated by heavy metals (i.e., Cu 2+ , Mn 2+ , Cr 3+ , Fe 3+ and Ni 2+ ) were highlighted. Kinetic modelling and microbial community dynamics were studied, according to each OLR, to carefully describe the UASG performance. The economic analysis showed a stable operation for the anaerobic digestion of unsaturated polyester resin wastewater using UASG, and the maximum net profit was achieved at OLR of 0.71 ± 0.08 gCOD/L/d.

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-15

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

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-12-15

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10{sup -3}) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the

The G4-ECONS Economic Evaluation Tool for Generation IV Reactor Systems and its Proposed Application to Deliberately Small Reactor Systems and Proposed New Nuclear Fuel Cycle Facilities. Annex IX

International Nuclear Information System (INIS)

2013-01-01

At the outset of the international Generation IV programme, it was decided that the six candidate reactor systems will ultimately be evaluated on the basis of safety, sustainability, non-proliferation attributes, technical readiness and projected economics. It is likely that the same factors will influence the evaluation of deliberately small reactor systems1 and new fuel cycle facilities, such as reprocessing plants that are being considered under the more recent Global Nuclear Energy Partnership (GNEP). This annex describes how the development of an economic modelling system has evolved to address the issue of economic competitiveness for both the Generation IV and GNEP programmes. In 2004, the Generation IV Economic Modelling Working Group (EMWG) commissioned the development of a Microsoft Excel based model capable of calculating the levelized unit electricity cost (LUEC) in mills/kW.h (1 mill = $10 -3 ) or $/MW.h for multiple types of reactor system being developed under the Generation IV programme. This overall modelling system is now called the Generation IV spreadsheet calculation of nuclear systems (G4-ECONS), and is being expanded to calculate costs of energy products in addition to electricity, such as hydrogen and desalinated water. A version has also been developed to evaluate the costs of products or services from fuel cycle facilities. The cost estimating methodology and algorithms are explained in detail in the Generation IV Cost Estimating Guidelines and in the G4-ECONS User's Manual. The model was constructed with relatively simple economic algorithms such that it could be used by almost any nation without regard to country specific taxation, cost accounting, depreciation or capital cost recovery methodologies. It was also designed with transparency to the user in mind (i.e. all algorithms and cell contents are visible to the user). A short description of version 1.0 G4-ECONS-R (reactor economics model) has also been published in the Proceedings of

Reactor safety in Eastern Europe

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

The Integral Fast Reactor

International Nuclear Information System (INIS)

Chang, Y.I.

1988-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

Factors affecting biological reduction of CO{sub 2} into CH{sub 4} using a hydrogenotrophic methanogen in a fixed bed reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Hyung; Pak, Daewon [Seoul National University of Science and Technology, Seoul (Korea, Republic of); Chang, Won Seok [Korea District Heating Corp, Seongnam (Korea, Republic of)

2015-10-15

Biological conversion of CO{sub 2} was examined in a fixed bed reactor inoculated with anaerobic mixed culture to investigate influencing factors, the type of packing material and the composition of the feeding gas mixture. During the operation of the fixed bed reactor by feeding the gas mixture (80% H{sub 2} and 20% CO{sub 2} based on volume basis), the volumetric CO{sub 2} conversion rate was higher in the fixed bed reactor packed with sponge due to its large surface area and high mass transfer from gas to liquid phase compared with PS ball. Carbon dioxide loaded into the fixed bed reactor was not completely converted because some of H{sub 2} was used for biomass growth. When a mole ratio of H{sub 2} to CO{sub 2} in the feeding gas mixture increased from 4 to 5, CO{sub 2} was completely converted into CH{sub 4}. The packing material with large surface area is effective in treating gaseous substrate such as CO{sub 2} and H{sub 2}. H{sub 2}, electron donor, should be providing more than required according to stoichiometry because some of it is used for biomass growth.

Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

Energy Technology Data Exchange (ETDEWEB)

Johnson, A.B. Jr.

1976-09-01

Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO/sub 2/) and lithium silicate (Li/sub 2/SiO/sub 3/) by the reaction: Li/sup 6/ + n ..-->.. /sup 4/He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100/sup 0/C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T/sub 2/), while in laboratory extractions (300-1300/sup 0/C), the tritium appeared primarily in the condensible form (HTO and T/sub 2/O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H/sub 2/O, CO/sub 2/, CO, O/sub 2/, H/sub 2/, NO, SO/sub 2/, SiF/sub 4/ and traces of hydrocarbons.

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

International Nuclear Information System (INIS)

Dupleac, D.; Perez, M.; Reventos, F.; Allison, C.

2011-01-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

Uncertainty analysis of the 35% reactor inlet header break in a CANDU 6 reactor using RELAP/SCDAPSIM/MOD4.0 with integrated uncertainty analysis option

Energy Technology Data Exchange (ETDEWEB)

Dupleac, D., E-mail: danieldu@cne.pub.ro [Politehnica Univ. of Bucharest (Romania); Perez, M.; Reventos, F., E-mail: marina.perez@upc.edu, E-mail: francesc.reventos@upc.edu [Technical Univ. of Catalonia (Spain); Allison, C., E-mail: iss@cableone.net [Innovative Systems Software (United States)

2011-07-01

The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology Software Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses the publicly available RELAP5 and SCDAP models in combination with (a) advanced programming and numerical techniques, (b) advanced SDTP-member-developed models for LWR, HWR, and research reactor analysis, and (c) a variety of other member-developed computational packages. One such computational package is an integrated uncertainty analysis (IUA) package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). RELAP/SCDAPSIM/MOD4.0(IUA) follows the input-propagation approach using probability distribution functions to define the uncertainty of the input parameters. The main steps for this type of methodologies, often referred as to statistical approaches or Wilks’ methods, are the ones that follow: 1. Selection of the plant; 2. Selection of the scenario; 3. Selection of the safety criteria; 4. Identification and ranking of the relevant phenomena based on the safety criteria; 5. Selection of the appropriate code parameters to represent those phenomena; 6. Association of uncertainty by means of Probability Distribution Functions (PDFs) for each selected parameter; 7. Random sampling of the selected parameters according to its PDF and performing multiple computer runs to obtain uncertainty bands with a certain percentile and confidence level; 8. Processing the results of the multiple computer runs to estimate the uncertainty bands for the computed quantities associated with the selected safety criteria. RELAP/SCDAPSIM/MOD4.0(IUA) calculates the number of required code runs given the desired percentile and confidence level, performs the sampling process for the

PERFORMA NEUTRONIK BAHAN BAKAR LiF-BeF2-ThF4-UF4 PADA SMALL MOBILE-MOLTEN SALT REACTOR

Directory of Open Access Journals (Sweden)

S. N. Rokhman

2015-04-01

Full Text Available Telah dilakukan analisis terhadap performa neutronik bahan bakar garam lebur LiF-BeF2-ThF4-UF4 pada Small Mobile-Molten Salt Reactor (SM-MSR. Penyesuaian konfigurasi teras dan temperatur operasi harus dilakukan untuk penggunaan bahan bakar baru tersebut agar mencapai keff > 1 dan CR (conversion ratio > 1 pada fraksi 0,5% 233U, 20% 232Th, 28% Li, 51,5% Be. Setelah didapat nilai keff ≈ 1 dan CR ≈ 1, dilakukan analisis pengaruh perubahan Th terhadap Be dan Be terhadap Li yang terlihat dalam perubahan parameter keff dan CR. Setelah itu fraksi 233U divariasi antara 0,5–0,46% untuk memperoleh keff > 1 dan CR > 1. Dalam perhitungan koefisien reaktifitas temperatur (αT, temperatur teras dinaikkan sebesar +25K dan +50K., dan untuk koefisien reaktifitas void (αV, densitas bahan bakar dikurangi hingga 90%. Hasil perhitungan menunjukkan bahwa pengurangan Th terhadap Be menyebabkan penurunan nilai CR dan naiknya keff akibat berkurangnya material fertil. Sebaliknya penambahan Be terhadap Li mengakibatkan terjadi kenaikan nilai keff dan menurunkan CR, akibat laju serapan Li lebih besar dari Be. Pada 5 (lima fraksi 233U dalam rentang 0,5–0,49%, hasil perhitungan keff dan CR masing-masing bervariasi dalam rentang 1,00001 - 1,00327 dan 1,00016 - 1,00731. Untuk faktor puncak daya (PPF, hasil perhitungan memberikan nilai dalam rentang 2,4311 -2,4714. Sedangkan untuk parameter keselamatan, koefisien reaktivitas temperatur (αT dan reaktivitas void (αV masingmasing bernilai negatif dalam rentang 4,972×10-5 - 5,909×10-5 dan 2,596×10-2- 2,8287×10-2 ∆k/k/K. Dapat disimpulkan bahwa teras SM-MSR memberikan nilai negatif di kedua koefisien reaktivitas tersebut untuk setiap fraksi,, sehingga memenuhi kriteria keselamatan dan keselamatan melekat. Kata kunci: SM-MSR (small mobile-molten salt reactor, bahan bakar LiF-BeF2-ThF4-UF4, keselamatan melekat, koefisien reaktivitas temperatur, koefisien reaktivitas void   The analysis of neutronic performance has

Spatial kinetics in nuclear reactor systems. Chapter 4

International Nuclear Information System (INIS)

Owens, D.H.

1980-01-01

The problem of constructing a low-order linear lumped-parameter model of xenon-induced spatial power oscillations in a large, cylindrical nuclear power reactor to replace an (assumed known) nonlinear distributed parameter model is examined. Model expansion and finite difference methods are used together to provide a successful solution to the problem. (U.K.)

Synthesis of superior fast charging-discharging nano-LiFePO4/C from nano-FePO4 generated using a confined area impinging jet reactor approach.

Science.gov (United States)

Liu, Xiao-min; Yan, Pen; Xie, Yin-Yin; Yang, Hui; Shen, Xiao-dong; Ma, Zi-Feng

2013-06-14

LiFePO4/C nanocomposites with excellent electrochemical performance is synthesized from nano-FePO4, generated by a novel method using a confined area impinging jet reactor (CIJR). When discharged at 80 C (13.6 Ag(-1)), the LiFePO4/C delivers a discharge capacity of 95 mA h g(-1), an energy density of 227 W h kg(-1) and a power density of 34 kW kg(-1).

Nuclear data usage for research reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

1996-01-01

In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

International Nuclear Information System (INIS)

Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

1991-09-01

The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

Proceedings of the 4th international symposium on material testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Suzuki, Masahide

2012-03-01

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

Proceedings of the 4th international symposium on material testing reactors

Energy Technology Data Exchange (ETDEWEB)

Ishihara, Masahiro; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

This report is the Proceedings of the fourth International Symposium on Material Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The first symposium was held on 2008, at the Oarai Research and Development Center of JAEA, the second, 2009, Idaho National Laboratory (INL) of United States and the third 2010, Nuclear Research Institute (NRI) in Czech Republic to exchange information for deep mutual understanding of material testing reactors. The fourth symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malasia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the general topics of 'status and future plan of material testing reactors', 'advancement of irradiation technology', 'expansion of industry use(RI)', 'facility, upgrade, aging management', 'new generation MTR', 'advancement of PIE technology', 'development of advanced driver fuel', and 'nuclear human resource development(HRD) for next generation', and 39 presentations were made. Furthermore, three topics, 'Necessity of cooperation for Mo-99 production by (n,gamma) reaction', 'Necessity of standardization of irradiation technology' and 'Conceptual design of next generation materials testing reactor by collaboration', were selected and discussed. (author)

Analysis of gamma dose for 4,8 gU/cm3 density silicide core at the RSG-GAS reactor using MCNP code

International Nuclear Information System (INIS)

Ardani

2011-01-01

Radiation safety analysis should be done following of substitution of fuel density of 2.96 gU/cc to density of 4,8 gU/cc silicide fuels for the RSG-GAS reactor. MCNP-5 code has been used to perform gamma dose calculation of the RSG-GAS reactor. Gamma radiation source at reactor consists of capture gamma rays, prompt fission gamma rays, and gamma rays of decay of fission and activation products. The strength of the prompt fission gamma rays is obtained by gamma releases of fission process of U-235 and reactor power of 30 MWt., during 46,6 days operation. Radiation dose is calculated at the experimental hall by detection point at the surface of outer of biological shielding and the operation hall by detection point at the top of the pool. The calculation is conducted at reactor on the normal operation and on the worst postulated accident causing the water level at the pool decreases. Calculation result shows that the biggest source strength of gamma rays come from the decay process. The highest calculated dose at the experiment hall is 4,07x10 -3 μSv/h, far from the maximum external dose permitted 25 μSv/h. The highest calculated dose at the operation hall is 19.98 μSv/h. Even though the calculated dose is still acceptable but this is close to the maximum permitted dose for worker. It concluded that loading of 4,8 gU/cc silicide fuel for the RSG-GAS still safe. (author)

Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

Directory of Open Access Journals (Sweden)

SA Hosseini

2013-09-01

Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation

Energy Technology Data Exchange (ETDEWEB)

Gomez-De Jesus, A.; Romano-Baez, F.J.; Leyva-Amezcua, L.; Juarez-Ramirez, C.; Ruiz-Ordaz, N. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico); Galindez-Mayer, J. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico)], E-mail: cmayer@encb.ipn.mx

2009-01-30

For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A{sub D}/A{sub R} ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values {epsilon}{sub G}, and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ({approx}95%), and by the stoichiometric release of chloride ions from the halogenated compound ({approx}80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions)

Radiation protection at the RA Reactor in 1993, RA research reactor, Part

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

1993-01-01

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

Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

2007-04-15

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

Heavy water moderated gas-cooled reactors

International Nuclear Information System (INIS)

Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

1964-01-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

Cermet fuel reactors

International Nuclear Information System (INIS)

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

Nuclear reactors to come

International Nuclear Information System (INIS)

Lung, M.

2002-01-01

The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

International Nuclear Information System (INIS)

Jammes, Ch.

2010-07-01

After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

Time to HAART Initiation after Diagnosis and Treatment of Opportunistic Infections in Patients with AIDS in Latin America.

Directory of Open Access Journals (Sweden)

Brenda Crabtree-Ramírez

Full Text Available Since 2009, earlier initiation of highly active antiretroviral therapy (HAART after an opportunistic infection (OI has been recommended based on lower risks of death and AIDS-related progression found in clinical trials. Delay in HAART initiation after OIs may be an important barrier for successful outcomes in patients with advanced disease. Timing of HAART initiation after an OI in "real life" settings in Latin America has not been evaluated.Patients in the Caribbean, Central and South America network for HIV Epidemiology (CCASAnet ≥18 years of age at enrolment, from 2001-2012 who had an OI before HAART initiation were included. Patients were divided in an early HAART (EH group (those initiating within 4 weeks of an OI and a delayed HAART (DH group (those initiating more than 4 weeks after an OI. All patients with an AIDS-defining OI were included. In patients with more than one OI the first event reported was considered. Calendar trends in the proportion of patients in the EH group (before and after 2009 were estimated by site and for the whole cohort. Factors associated with EH were estimated using multivariable logistic regression models.A total of 1457 patients had an OI before HAART initiation and were included in the analysis: 213 from Argentina, 686 from Brazil, 283 from Chile, 119 from Honduras and 156 from Mexico. Most prevalent OI were Tuberculosis (31%, followed by Pneumocystis pneumonia (24%, Invasive Candidiasis (16% and Toxoplasmosis (9%. Median time from OI to HAART initiation decreased significantly from 5.7 (interquartile range [IQR] 2.8-12.1 weeks before 2009 to 4.3 (IQR 2.0-7.1 after 2009 (p<0.01. Factors associated with starting HAART within 4 weeks of OI diagnosis were lower CD4 count at enrolment (p-<0.001, having a non-tuberculosis OI (p<0.001, study site (p<0.001, and more recent years of OI diagnosis (p<0.001.The time from diagnosis of an OI to HAART initiation has decreased in Latin America coinciding with the

Children with severe Osteogenesis imperfecta and short stature present on average with normal IGF-I and IGFBP-3 levels.

Science.gov (United States)

Hoyer-Kuhn, Heike; Höbing, Laura; Cassens, Julia; Schoenau, Eckhard; Semler, Oliver

2016-07-01

Osteogenesis imperfecta (OI) is characterized by bone fragility and short stature. Data about IGF-I/IGFBP-3 levels are rare in OI. Therefore IGF-I/IGFBP-3 levels in children with different types of OI were investigated. IGF-I and IGFBP-3 levels of 60 children (male n=38) were assessed in a retrospective cross-sectional setting. Height/weight was significant different [height z-score type 3 versus type 4: p=0.0011 and weight (p≤0.0001)] between OI type 3 and 4. Mean IGF-I levels were in the lower normal range (mean±SD level 137.4±109.1 μg/L). Mean IGFBP-3 measurements were in the normal range (mean±SD 3.105±1.175 mg/L). No significant differences between OI type 3 and 4 children have been observed (IGF-I: p=0.0906; IGFBP-3: p=0.2042). Patients with different severities of OI have IGF-I and IGFBP-3 levels in the lower normal range. The type of OI does not significantly influence these growth factors.

The integral fast reactor concept

International Nuclear Information System (INIS)

Chang, Yoon I.; Marchaterre, J.F.

1987-01-01

The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

Current scenario of opportunistic and co-infections in HIV-infected individuals at a tertiary care hospital in Mumbai, India.

Science.gov (United States)

Chavan, V R; Chaudhary, V; Ahir, P; Mehta, R; Mavani, P S; Kerkar, C; Pramanik, J M

2015-01-01

An update on opportunistic infections/co-infections (OIs/CIs) is essential to understand the success of highly active antiretroviral therapy offered by the government agencies in reducing AIDS-related OIs/CIs. Hence, the present study aimed to evaluate the frequency of OIs/CIs in HIV-positive individuals at a tertiary care hospital in Mumbai. Its' association with CD4 counts, anti-retroviral treatment and on HIV transmission was also determined. An observational study was designed to evaluate different OIs/CIs in individuals, who tested positive for HIV infection at the ICTC/Shakti Clinic of Seth G.S. Medical College and KEM Hospital, Mumbai. Data analysis was done with the use of SPSS software (version 19.0, SPSS, Chicago, IL, USA). P value was considered significant if it is < 0.05. Heterosexual contact was the major route of transmission among the enrolled 185 individuals. Ninety (48.06%) HIV-infected individuals were with OIs/CIs. Tuberculosis (TB) was the most common OI (68.8%). Other CIs noted were Herpes zoster, syphilis, hepatitis C and B, malaria, typhoid and dengue. The median CD4 count in HIV-positive individuals with TB was 337 ± 248 cells/μl, and 67.7% of individuals with OIs/CIs had low CD4 counts (<400 cells/μl). Individuals in 31-40 years of age group had significantly (P = 0.01) more OIs/CIs. More (53.7%) spouse/children of HIV-positive individuals without OIs/CIs were HIV-1 positive. Low proportions of individuals with or without OIs/CIs were on ART. Nearly half of HIV-infected individuals were with OIs/CIs. Initiation of free ART programme since 2004 possibly associated with the type and rate of OIs/CIs. Tuberculosis and multiple OIs/CIs were associated with low CD4 counts. Infection was high in 31-40 years age group. Most of the spouses of individuals without OIs/CIs were HIV positive, indirectly indicates lack of condom use or lack of awareness of condom use.

Reactor safety in Eastern Europe. Proceedings

International Nuclear Information System (INIS)

1995-02-01

The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. (HP) [de

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-15

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

Mirror hybrid reactor optimization studies

International Nuclear Information System (INIS)

Bender, D.J.

1976-01-01

A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

Neutrino-4 experiment on the search for a sterile neutrino at the SM-3 reactor

Energy Technology Data Exchange (ETDEWEB)

Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Chernyi, A. V.; Zherebtsov, O. M. [National Research Centre “Kurchatov Institute,”, Konstantinov Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I. [National Research Centre “Kurchatov Institute,” (Russian Federation); Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K. [State Scientific Centre Research Institute of Atomic Reactors (Russian Federation); and others

2015-10-15

In view of the possibility of the existence of a sterile neutrino, test measurements of the dependence of the reactor antineutrino flux on the distance from the reactor core has been performed on SM-2 reactor with the Neutrino-2 detector model in the range of 6–11 m. Prospects of the search for reactor antineutrinos at short distances have been discussed.

Speech to be delivered by Mr. François de Rose, president of Council of the european organization for nuclear research on the occasion of the inauguration of the CERN proton synchrotron on 5 february 1960

CERN Multimedia

CERN Press Office. Geneva

1960-01-01

Speech to be delivered by Mr. François de Rose, president of Council of the european organization for nuclear research on the occasion of the inauguration of the CERN proton synchrotron on 5 february 1960

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Cupac, S.; Stanic, A.

1990-01-01

RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

Reactor core structure

International Nuclear Information System (INIS)

Higashinakagawa, Emiko; Sato, Kanemitsu.

1992-01-01

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

Molten salt reactors

International Nuclear Information System (INIS)

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

2014-01-01

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

Possibilities of TWR and long life reactor

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Shimazu, Yoichiro; Handa, Norihiko

2010-01-01

Bill Gates identified the need to switch to zero-emission energy and clarified investing in Terra Power developing the TWR (Traveling Wave Reactor) in February 2010. He also visited Toshiba developing small reactor 4S (Super Safe Small and Simple). In Japan design studies of the TWR have been conducted on the CANDLE reactor without refueling and the 4S long life reactor with maintenance free. In this feature article, the state of R and D on the TWR in Japan and IAEA's activities on small reactors without online refueling were reviewed in addition to articles on impacts of Bill Gates' investment in the TWR and state of the TWR development from an interview with John Gilleland of Terra Power. (T. Tanaka)

Public information circular for shipments of irradiated reactor fuel. Revision 4

International Nuclear Information System (INIS)

1984-06-01

This publication is the fifth in a series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. This publication contains basically three kinds of information: (1) routes recently approved (18 months) by the Commission for the shipment of irradiated reactor fuel; (2) information regarding any safeguards-significant incidents that may be (to date none have) reported during shipments along such routes; and (3) cumulative amounts of material shipped

Status of French reactors

Energy Technology Data Exchange (ETDEWEB)

Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

1982-12-01

Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2014. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Osa, Akihiko; Imahashi, Masaki; Hirane, Nobuhiko; Motome, Yuiko; Tayama, Hidekazu; Tamura, Itaru; Harada, Yuko; Sakata, Mami; Kadokura, Masakazu; Takita, Chiharu

2017-02-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3 (Japan Research Reactor No.3), JRR-4 (Japan Research Reactor No.4), NSRR (Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2014. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration, and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2013. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Kashima, Yoichi; Murayama, Yoji; Nakamura, Kiyoshi; Uno, Yuki; Hirane, Nobuhiko; Ohuchi, Hitoshi; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Harada, Yuko; Kadokura, Masakazu; Machi, Sumire; Takita, Chiharu

2015-02-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2013. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

CO2 photoreduction using NiO/InTaO4 in optical-fiber reactor for renewable energy

NARCIS (Netherlands)

Wang, Zhen-Yi; Chou, Hung-Chi; Wu, Jeffrey C.S.; Tsai, Din Ping; Mul, Guido

2010-01-01

The photocatalytic reduction of CO2 into fuels provides a direct route to produce renewable energy from sunlight. NiO loaded InTaO4 photocatalyst was prepared by a sol–gel method. Aqueous-phase CO2 photoreduction was performed in a quartz reactor to search for the highest photoactivity in a series

Kinetic analysis of sub-prompt-critical reactor assemblies

International Nuclear Information System (INIS)

Das, S.

1992-01-01

Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

Sodium cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Hokkyo, N; Inoue, K; Maeda, H

1968-11-21

In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

Mirror reactor studies

International Nuclear Information System (INIS)

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

1977-01-01

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

Low Enrichment Uranium (LEU)-fueled SLOWPOKE-2 nuclear reactor simulation with the Monte-Carlo based MCNP 4A code

International Nuclear Information System (INIS)

Pierre, J.R.M.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.

Turning points in reactor design

International Nuclear Information System (INIS)

Beckjord, E.S.

1995-01-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

Turning points in reactor design

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-09-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

Aqueous homogeneous suspension reactor project. Report over the 4th quarter and the year 1974

Energy Technology Data Exchange (ETDEWEB)

1975-07-01

The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described.

Introduction of advanced pressurized water reactors in France

International Nuclear Information System (INIS)

Millot, J.P.; Nigon, M.; Vitton, M.

1988-01-01

Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Fast reactor irradiation effects on fracture toughness of Si_3N_4 in comparison with MgAl_2O_4 and yttria stabilized ZrO_2

International Nuclear Information System (INIS)

Tada, K.; Watanabe, M.; Tachi, Y.; Kurishita, H.; Nagata, S.; Shikama, T.

2016-01-01

Fracture toughness of silicon nitride (Si_3N_4), magnesia-alumina spinel (MgAl_2O_4) and yttria stabilized zirconia (8 mol%Y_2O_3–ZrO_2) was evaluated by the Vickers-indentation technique after the fast reactor irradiation up to 55 dpa (displacement per atom) at about 700 °C in the Joyo. The change of the fracture toughness by the irradiation was correlated with nanostructural evolution by the irradiation, which was examined by transmission electron microscopy. The observed degradation of fracture toughness in Si_3N_4 is thought to be due to the relatively high density of small-sized of the irradiation induced defects, which should be resulted from a large amount of transmutation gases of hydrogen and helium. Observed improvement of fracture toughness in MgAl_2O_4 was due to the blocking of crack propagation by the antiphase boundaries. The radiation effects affected the fracture toughness of yttria stabilized zirconia at 55 dpa, suggesting that the generated high density voids would affect the propagation of cracks. - Highlights: • Si_3N_4, MgAl_2O_4 and YSZ were neutron irradiated up to 55dpa around 700 °C in the Joyo. • They are candidate ceramics for the inert matrices of nuclear fuels in the fast reactors. • The irradiation enhanced the fracture toughness of MgAl_2O_4 and YSZ, while degraded that of Si_3N_4. • The toughness changes were correlated with radiation induced defects and transmutation gases.

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

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

Reactor Physics Behind the Chernobyl Accident

International Nuclear Information System (INIS)

Reisch, F.

1999-01-01

There are some fourteen Chernobyl type of power reactors (1000 MWe) in operation at five different sites in Eastern Europe. In Russia; in St. Petersburg (4). in Smolensk (3). and in Kursk (4) in the Ukraine in Chernobyl (l) and in Lithuania in Ignalina (2). The oldest one is west of St. Petersburg and the most powerful one is in Ignalina. The reactors at St. Petersburg and in Lithuania are near to the Baltic sea. An intricate reactor construction was the most important cause of the accident. There were other reasons too: human error. politics and economics

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

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

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

Office of Nuclear Regulatory Research summary of advanced reactors activities, June 4, 2001

International Nuclear Information System (INIS)

2001-01-01

Pre-application interactions with potential licensee applicants will help NRC prepare for future submittals, through the development of the infrastructure necessary for licensing application reviews. RES has the lead for non-LWR advanced reactor pre-application initiatives and longer-range new technology initiatives. An advanced reactor group has been formed in REAHFB, and is currently performing a pre-application review of Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre application interaction include General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) design. RES advanced reactors activities also include participation as an observer in DOE's Generation IV initiative. Pre-Application review objectives include the development of regulatory guidance, licensing approach, and technology-basis expectations for licensing advanced designs, including identifying significant technology, design, safety, licensing and policy issues that would need to be addressed in the licensing process. The presentation described the pre-application process for the Exelon PBMR. NRC first identifies additional information following topical meetings with Exelon, and Exelon formally documents and submits required topical Information. The staff then develops a preliminary assessment and drafts a response which is followed by stakeholder input and comments at a public workshop. Preliminary assessments are discussed with ACRS and ACNW, and Commission papers are written which provide staff positions and recommendations on proposed policy decisions. Some of the significant areas for the PBMR include: Process Issues, Legal and Financial Issues; Regulatory Framework; Fuel Performance and Qualification; Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials); Fuel Cycle Safety Areas; PRA, SSC Safety Classification; PBMR Prototype Testing

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

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

1989-01-01

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

Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

DEFF Research Database (Denmark)

Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

2013-01-01

Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.; Miller, J.R.

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1973-01-01

The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

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

2011-08-15

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

Astrid (fast breeder nuclear reactor)

International Nuclear Information System (INIS)

2014-01-01

This document presents ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a French project of sodium-cooled fast breeder reactor, fourth generation reactor which should be fuelled by uranium 238 rather than uranium 235, and should therefore need less extracted natural uranium to produce electricity. The operation principle of fast breeder reactors is described. They notably directly consume plutonium, allow an easier radioactive waste management as they transform long life radioactive elements into shorter life elements by transmutation. The regeneration process is briefly described, and the various operation modes are evoked (iso-generator, sub-generator, and breeder). Some peculiarities of sodium-cooled reactors are outlined. The Astrid operation principle is described, its main design innovations outlined. Various challenges are discussed regarding safety of supply and waste processing, and the safety of future reactors. Major actors are indicated: CEA, Areva, EDF, SEIV Alcen, Toshiba, Rolls Royce, and Comex. Some key data are indicated: expected lifetime, expected availability rate, cost. The projected site is Marcoule and fast breeder reactors operated or under construction in the world are indicated. The document also proposes an overview of the background and evolution of reactors of 4. generation

Reactor safety analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

Improvement of research reactor sustainability

International Nuclear Information System (INIS)

Ciocanescu, M.; Paunoiu, C.; Toma, C.; Preda, M.; Ionila, M.

2010-01-01

The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)

Factors associated with development of opportunistic infections in HIV-1 infected adults with high CD4 cell counts: a EuroSIDA study

DEFF Research Database (Denmark)

Podlekareva, Daria; Mocroft, A; Dragsted, Ulrik Bak

2006-01-01

BACKGROUND: Limited data exist on factors predicting the development of opportunistic infections (OIs) at higher-than-expected CD4(+) cell counts in human immunodeficiency virus (HIV) type 1-infected adults. METHODS: Multivariate Poisson regression models were used to determine factors related to...

Fukushima - calculation of the reactor core inventory and storage pools Dai-ichi 1 to Dai-ichi 4, an estimation of a source term

International Nuclear Information System (INIS)

Krpelanova, M.; Carny, P.

2011-01-01

Inventory of the reactor core and spent fuel storage pool of the reactors at Dai-ichi 1 to Dai-ichi 4 was determined to need a realistic estimate of the source (released into the atmosphere environment) and modelling of radiological impact of the events in Fukushima NPP. Calculations of inventories were carried out by the methodology that is used in systems to support emergency response and crisis management anymore. Calculations were made based on a model that respects knowledge of real fuels and fuel cycles for individual reactors Dai-ichi. Necessary input data for training the model and calculate inventories are obtained from the IAEA PRIS database.

Safety philosophy and safety technology of the Soviet RBMK reactors

International Nuclear Information System (INIS)

Zuend, H.; Jarvis, A.S.; Haennis, H.P.; Tikal, J.

1986-01-01

Safety requirements and control in USSR are outlined. Safety criteria and practical application in the case of the RBMK type reactor Chernobyl-4 are discussed. An overview of the Chernobyl-4 reactor accident including its causes is given. Measures to improve the safety of RBMK reactors are described

X-Ray Brightening and UV Fading of Tidal Disruption Event ASASSN-15oi

Science.gov (United States)

Gezari, S.; Cenko, S. B.; Arcavi, I.

2017-12-01

We present late-time observations by Swift and XMM-Newton of the tidal disruption event (TDE) ASASSN-15oi that reveal that the source brightened in the X-rays by a factor of ∼10 one year after its discovery, while it faded in the UV/optical by a factor of ∼100. The XMM-Newton observations measure a soft X-ray blackbody component with {{kT}}{bb}∼ 45 {eV}, corresponding to radiation from several gravitational radii of a central ∼ {10}6 {M}ȯ black hole. The last Swift epoch taken almost 600 days after discovery shows that the X-ray source has faded back to its levels during the UV/optical peak. The timescale of the X-ray brightening suggests that the X-ray emission could be coming from delayed accretion through a newly forming debris disk and that the prompt UV/optical emission is from the prior circularization of the disk through stream–stream collisions. The lack of spectral evolution during the X-ray brightening disfavors ionization breakout of a TDE “veiled” by obscuring material. This is the first time a TDE has been shown to have a delayed peak in soft X-rays relative to the UV/optical peak, which may be the first clear signature of the real-time assembly of a nascent accretion disk, and provides strong evidence for the origin of the UV/optical emission from circularization, as opposed to reprocessed emission of accretion radiation.

Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

International Nuclear Information System (INIS)

Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

2007-01-01

Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

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

BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup

International Nuclear Information System (INIS)

1998-01-01

1 - Description of program or function: The system of codes can be used to solve nuclear reactor core static neutronics and reactor history exposure problems. BOLD/VENTURE-4: First order perturbation and time-dependent sensitivity theories can be applied. Control rod positioning may be modeled explicitly and refueling treated with repositioning and recycle. Special capability is coded to model the continuously fueled core and to solve the importance and dominant harmonics problems. The modules of the code system are: VENTNEUT: VENTURE neutronics module; DRIVER and CONTRL: Control module; BURNER: Exposure calculation for reactor core analysis; FILEDTOR: File editor; INPROSER: Input processor; EXPOSURE: BURNER code module; REACRATE: Reaction rate calculation; CNTRODPO: Control rod positioning; FUELMANG: Fuel management positioning and accounting; PERTUBAT: Perturbation reactivity importance analyses; sensitivity analysis; DEPTHMOD: Static and time-dependent perturbation sensitivity analysis. The special processors are: DVENTR: Handles the input to the VENTURE module; DCMACR: Converts CITATION macroscopic cross sections to microscopic cross sections; DCRSPR: Produces input for the CROSPROS module; DUTLIN: Adds or replaces problem input data without exiting the program; DENMAN: Repositions fuel; DMISLY: Miscellaneous tasks. Standard interface files between modules are binary sequential files that follow a standardized format. VENTURE-PC: The microcomputer version is a subset of the mainframe version. The modules and special processors which are not part of VENTURE-PC are: REACRATE, CNTRODPO, PERTUBAT, FUELMANG, DEPTHMOD, DMISLY. 2 - method of solution: BOLD-VENTURE-4: The neutronics problems are solved by applying the multigroup diffusion theory representation of neutron transport applying an over-relaxation inner iteration, outer iteration scheme. Special modeling is used or source correction is done during iteration to solve importance and harmonics problems. No

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-31

The chapter briefly discussed the following subjects: radiological protection problems of reactor 1. in operation 2. types of reactor i.e. power reactors, research reactors, etc. 3. during maintenance and installation of fuels. 4. nuclear fuels.

Development of a helical-coil double wall tube steam generator for 4S reactor

International Nuclear Information System (INIS)

Kitajima, Yuko; Maruyama, Shigeki; Jimbo, Noboru; Hino, Takehisa; Sato, Katsuhiko

2011-01-01

The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water stream systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref. 1) and Ref. 2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG. (author)

Contribution to the improvement of the evaluation methods of nuclear heating in reactors by using the Monte Carlo code TRIPOLI-4

International Nuclear Information System (INIS)

Peron, Arthur

2014-01-01

Technological irradiation programs carried out in experimental reactors are crucial for the support of the current nuclear fleet in terms of study and anticipation of the behavior under irradiation of fuels and structural materials. These programs make it possible to improve the safety of the current reactors and also to study materials for the new concepts of reactors. Irradiation conditions of materials in experimental reactors must be representative of those of nuclear power plants (NPPs). One of the main advantages of material testing reactors (MTRs) is to be able to carry out instrumented irradiations by adjusting experimental parameters, in particular the neutron flux and the temperature. The control of the parameter temperature of a device irradiated in an experimental reactor requires the knowledge of the nuclear heating (source term) due to the deposition of energy of the photons and the neutrons interacting in the device. A relevant evaluation of this heating is a key data for the thermal studies of design and safety of devices. The objective of this thesis is to improve the methods of the evaluation of nuclear heating in reactors. This work consists of the development of an innovating and complete coupled neutron-photon calculation scheme (allowing to obtain the contribution of neutrons, prompt gamma and decay gamma), mainly based on the 3D, continuous energy TRIPOLI-4 Monte Carlo transport code. An experimental validation of the calculation scheme has been performed, based on calorimetry measurements carried out in the OSIRIS reactor at CEA Saclay. Sensitivity studies have been undertaken to establish the impact of various parameters on nuclear heating calculations (in particular nuclear data) and to fix the final calculation scheme to be closer to the technological irradiation aspects. The thesis work leads to an operational and predictive tool for the nuclear heating estimation, meeting the experimentation needs of research reactors and can be

Corrosion of reactor materials

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-01-15

Much operational experience and many experimental results have accumulated in recent years regarding corrosion of reactor materials, particularly since the 1958 Geneva Conference on the Peaceful Uses of Atomic Energy, where these problems were also discussed. It was, felt that a survey and critical appraisal of the results obtained during this period had become necessary and, in response to this need, IAEA organized a Conference on the Corrosion of Reactor Materials at Salzburg, Austria (4-9 June 1962). It covered many of the theoretical, experimental and engineering problems relating to the corrosion phenomena which occur in nuclear reactors as well as in the adjacent circuits

Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

Magnetite nanoparticles enhance the performance of a combined bioelectrode-UASB reactor for reductive transformation of 2,4-dichloronitrobenzene.

Science.gov (United States)

Wang, Caiqin; Ye, Lu; Jin, Jie; Chen, Hui; Xu, Xiangyang; Zhu, Liang

2017-09-04

Direct interspecies electron transfer (DIET) among the cometabolism microbes plays a key role in the anaerobic degradation of persistent organic pollutants and stability of anaerobic bioreactor. In this study, the COD removal efficiency increased to 99.0% during the start-up stage in the combined bioelectrode-UASB system (R1) with magnetite nanoparticles addition, which was higher than those in the coupled bioelectrode-UASB (R2; 83.2%) and regular UASB (R3; 71.0%). During the stable stage, the increase of 2,4-dichloronitrobenzene (2,4-DClNB) concentration from 25 mg L -1 to 200 mg L -1 did not affect the COD removal efficiencies in R1 and R2, whereas the performance of R3 was deteriorated obviously. Further intermediates analysis indicated that magnetite nanoparticles enhanced the reductive dechlorination of 2,4-DClNB. High-throughput sequencing results showed that the functional microbes like Syntrophobacter and Syntrophomonas which have been reported to favor the DIET, were predominant on the cathode surface of R1 reactor. It is speculated that the addition of magnetite nanoparticles favors the cooperative metabolism of dechlorinating microbes and electricigens during 2,4-DClNB degradation process in the combined bioelectrode-UASB reactor. This study may provide a new strategy to improve the performance of microbial electrolysis cells and enhance the pollutant removal efficiency.

Development of an innovative reflector drive mechanism using magnetic repulsion force for 4S reactor

International Nuclear Information System (INIS)

Tsuji, K.; Watanabe, M.; Inagaki, H.; Nishikawa, A.; Takahashi, H.; Wakamatsu, M.; Matsumiya, H.; Nishiguchi, Y.

2001-01-01

A small sized fast reactor 4S: (Super Safe Small and Simple) which has a core of 10 - 30 years life time is controlled by reflectors. The reflector is required to be risen at very low speed to make up for the reactivity swing during operation. This report shows the development of an innovative reflector drive mechanism using magnetic repulsion force that can move at a several micrometer per one step. This drive mechanism has a passive shut down capability, and can eliminate reflector drive line. (author)

Nuclear Effects Analysis D1-S-1800 Aerial Radiac System AN/ADR-6(XE-4) (V).

Science.gov (United States)

1972-12-21

1 ,r-4»-!4.. l.F-4,l<»..l .«14. TAP |. FT = 0.n.c;,OF-A.n.c.oiF-nft . 1 . OF 8 « 7 . 4<^P-* , 1 . OF« . 7 . SF-6 • 0 * 1 .F-2,0 TAPI[ F4...CffC49\\A|f r* if p- or < ( eft R <\\r\\r^f^OftCC~~ ’v.fvtvrv.fv — rv^. •* i i i i I i njjliflf. -ceo cccoccccc I I I I I I

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

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

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

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

Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

2013-10-15

Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-07-01

A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-09-01

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

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

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

Science.gov (United States)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

Energy Technology Data Exchange (ETDEWEB)

Davenport, Michael; Petti, D. A.

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

Fusion reactor materials

International Nuclear Information System (INIS)

Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

1980-01-01

Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

Fusion reactors as a future energy source

International Nuclear Information System (INIS)

Seifritz, W.

A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

How many reactor accidents will there be

International Nuclear Information System (INIS)

Islam, S.; Lindgren, K.

1986-01-01

A method for calculation of the probability of nuclear accidents is described. The method is based on the use of data from reactor operating experience, i.e. there have been two major accidents [Three Mile Island and Chernobyl] during 4,000 reactor-years (cumulative operating experience). The authors argue that this method is better than the present ''technical risk assessment'' method based on the likelihood of failure of a reactor component or safety system, used by designers of nuclear reactor. (U.K.)

Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

DEFF Research Database (Denmark)

Ippolito, Davide; Kammer Hansen, Kent

2014-01-01

The reduction of air pollution has become an international concern over the last ten years because of increases in emissions from mobile and stationary sources. Among these sources, volatile organic compounds (VOC) represent a serious environmental problem, together with NOx, SOx and particulate...... VOC component of Diesel engine exhausts, over a wide range of temperatures. The entire reactor was thought as a highly porous catalytic filter for a possible application in a Diesel exhausts purification system. The porous reactor was used as a backbone for the infiltration of Co3O4 and Co3O4/CeO2...

GHRSST Level 4 DMI_OI Global Foundation Sea Surface Temperature Analysis (GDS version 2)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — A Group for High Resolution Sea Surface Temperature (GHRSST) Level 4 sea surface temperature analysis produced daily on an operational basis by the Danish...

Spectrums of opportunistic infections and malignancies in HIV-infected patients in tertiary care hospital, China.

Directory of Open Access Journals (Sweden)

Jiang Xiao

Full Text Available BACKGROUND: HIV-related opportunistic infections (OIs and malignancies continued to cause morbidity and mortality in Chinese HIV-infected individuals. The objective for this study is to elucidate the prevalence and spectrums of OIs and malignancies in HIV-infected patients in the Beijing Ditan Hospital. METHODS: The evaluation of the prevalence and spectrums of OIs and malignancies was conducted by using the clinical data of 834 HIV-infected patients admitted in the Beijing Ditan hospital from January 1, 2009, to November 30, 2012. RESULTS: The prevalence and spectrums of OIs and malignancies varied contingent on geographic region, transmission routes, and CD4 levels. We found that tuberculosis was most common OI and prevalence was 32.5%, followed by candidiasis(29.3%, Pneumocystis pneumonia(PCP(22.4%, cytomegalovirus(CMV infection(21.7%, other fungal infections(16.2%, mycobacterium avium complex(MAC(11.3%, cryptococcosis(8.0%, progressive multifocal leukoencephalopathy(PML(4.4%, Cerebral Toxoplasmosis(3.5% and Penicillium marneffei infection(1.4%; while Lymphoma(2.9%, Kaposi's sarcoma(0.8% and cervix carcinoma(0.3% were emerged as common AIDS-defining malignancies. Pulmonary OI infections were the most prevalent morbidity and mortality in patients in the AIDS stage including pulmonary tuberculosis (26.6% and PCP (22.4%. CMV infection(21.7% was most common viral infection; Fungal OIs were one of most prevalent morbidity in patients in the AIDS stage, including oral candidiasis (29.3%, other fungal infection (16.2%, Cryptococcosis (8.0% and Penicillium marneffei infection (1.4%. We found the low prevalence of AIDS-defining illnesses in central neural system in this study, including progressive multifocal leukoencephalopathy (4.4%, cerebral toxoplasmosis (3.5%, tuberculosis meningitis (3.2%, cryptococcal meningitis (2.4% and CMV encephalitis (1.1%. In-hospital mortality rate was 4.3 per 100 person-years due to severe OIs, malignancies, and medical

Research and development of a super fast reactor (12). Considerations for the reactor characteristics

International Nuclear Information System (INIS)

Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

2008-01-01

A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

Deposition of RuO 4 on various surfaces in a nuclear reactor containment

Science.gov (United States)

Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

2009-07-01

During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

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

1987-01-01

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

Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

2006-11-15

This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

GHRSST Level 4 MW_OI Global Foundation Sea Surface Temperature analysis (GDS version 2)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — A Group for High Resolution Sea Surface Temperature (GHRSST) global Level 4 sea surface temperature analysis produced daily on a 0.25 degree grid at Remote Sensing...

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

2016-09-26

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

Directory of Open Access Journals (Sweden)

Guidez Joel

2017-01-01

In the case of sodium-cooled fast reactors (SFRs, the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction. From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

Inter renewal travelling wave reactor with rotary fuel columns

International Nuclear Information System (INIS)

Terai, Yuzo

2016-01-01

To realize the COP21 decision, this paper proposes Inter Renewal Travelling Wave Reactor that bear high burn-up rate 50% and product TRU fuel efficiently. The reactor is based on 4S Fast Reactor and has Reactor Fuel Columns as fuel assemblies that equalize temperature in the fuel assembly so that fewer structure is need to restrain thermal transformation. To equalize burn-up rate of all fuel assemblies in the reactor, each rotary fuel column has each motor-lifter. The rotary fuel column has two types (Cylinder type and Heat Pipe type using natrium at 15 kPa which supply high temperature energy for Ultra Super Critical power plant). At 4 years cycle all rotary fuel columns of the reactor are renewed by the metallurgy method (vacuum re-smelting) and TRU fuel is gotten from the water fuel. (author)

BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

International Nuclear Information System (INIS)

1987-11-01

This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

Overview of fusion reactor safety

International Nuclear Information System (INIS)

Cohen, S.; Crocker, J.G.

1981-01-01

Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

An efficient RuCl3·H2O/I2 catalytic system: A facile access to 3-aroylimidazo[1,2-a]pyridines from 2-aminopyridines and chalcones

Directory of Open Access Journals (Sweden)

P.V. Sri Ramya

2018-01-01

Full Text Available A simple and efficient protocol has been demonstrated for the preparation of densely functionalized 3-aroylimidazo[1,2-a]pyridines from 2-aminopyridines and chalcones using RuCl3·H2O/I2 catalytic system. The advantages, such as low catalyst loading, broad substrate scope with respect to substitutions on aminopyridines as well as chalcones, stability of heterocycles such as thiophene under the reaction conditions, operationally simple procedure and higher yields makes this approach remarkable for synthetic applications.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

International Nuclear Information System (INIS)

Van Swam, L.F.P.; Shann, S.H.

1991-01-01

This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

Isothermal calorimeter for reactor radiation dosimetry

Energy Technology Data Exchange (ETDEWEB)

Radak, B; Markovic, V [Institute of Nuclear Sciences Boris Kidric, Odeljenje za radijacionu hemiju, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

An isothermal calorimeter with thermistors for measuring absorbed dose rates from 10{sup 4}-5-6.10{sup 5} rad/h in reactor experimental holes has been designed. A kinetics method for determining the equilibrium temperature difference has been developed, and its application in isothermal calorimetry proved. The expected accuracy in measurements within {+-} 2-5% has been proved by measurements carried out in the reactor. Some data obtained by measurements in the reactor RA are presented (author)

Advanced designs of VVER reactor plant

International Nuclear Information System (INIS)

Mokhov, V.A.

2010-01-01

The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

François Le Guévellou, L'ésotérisme et les sciences occultes en Russie. Aperçu historique et lexiques.

Directory of Open Access Journals (Sweden)

Olivier Santamaria

2009-06-01

Full Text Available Professeur de linguistique russe à l'INALCO, François Le Guévellou s'est fait connaître du public spécialisé par la rédaction de dictionnaires français-russe / russe-français présentant des lexiques quelque peu originaux (Dictionnaire des gros mots russes ; Dictionnaire russe-français des noms d’animaux et de plantes ; Les termes de couleurs en français et en russe. Etude contrastive et lexiques ; Dictionnaire des onomatopées et interjections russes, souvent accompagnés de savantes études in...

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Proceedings of the advisory committee on reactor safeguards workshop on future reactors

International Nuclear Information System (INIS)

2001-12-01

This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda

Proceedings of the advisory committee on reactor safeguards workshop on future reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-12-01

This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda.

Rapid preparation of high electrochemical performance LiFePO4/C composite cathode material with an ultrasonic-intensified micro-impinging jetting reactor.

Science.gov (United States)

Dong, Bin; Huang, Xiani; Yang, Xiaogang; Li, Guang; Xia, Lan; Chen, George

2017-11-01

A joint chemical reactor system referred to as an ultrasonic-intensified micro-impinging jetting reactor (UIJR), which possesses the feature of fast micro-mixing, was proposed and has been employed for rapid preparation of FePO 4 particles that are amalgamated by nanoscale primary crystals. As one of the important precursors for the fabrication of lithium iron phosphate cathode, the properties of FePO 4 nano particles significantly affect the performance of the lithium iron phosphate cathode. Thus, the effects of joint use of impinging stream and ultrasonic irradiation on the formation of mesoporous structure of FePO 4 nano precursor particles and the electrochemical properties of amalgamated LiFePO 4 /C have been investigated. Additionally, the effects of the reactant concentration (C=0.5, 1.0 and 1.5molL -1 ), and volumetric flow rate (V=17.15, 51.44, and 85.74mLmin -1 ) on synthesis of FePO 4 ·2H 2 O nucleus have been studied when the impinging jetting reactor (IJR) and UIJR are to operate in nonsubmerged mode. It was affirmed from the experiments that the FePO 4 nano precursor particles prepared using UIJR have well-formed mesoporous structures with the primary crystal size of 44.6nm, an average pore size of 15.2nm, and a specific surface area of 134.54m 2 g -1 when the reactant concentration and volumetric flow rate are 1.0molL -1 and 85.74mLmin -1 respectively. The amalgamated LiFePO 4 /C composites can deliver good electrochemical performance with discharge capacities of 156.7mAhg -1 at 0.1C, and exhibit 138.0mAhg -1 after 100 cycles at 0.5C, which is 95.3% of the initial discharge capacity. Copyright © 2017. Published by Elsevier B.V.

A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

International Nuclear Information System (INIS)

Limaiem, I.

2006-12-01

Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

International Nuclear Information System (INIS)

Gustavsson, Veine

2002-11-01

The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

Calculation of the power distribution in the fuel rods of the low power research reactor using the MCNP4C code

International Nuclear Information System (INIS)

Dawahra, S.; Khattab, K.

2011-01-01

Highlights: → The MCNP4C code was used to calculate the power distribution in 3-D geometry in the MNSR reactor. → The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. → The minimum power was found in the fuel ring number 9 and was 79.9 W. → The total power in the total fuel rods was 30.9 kW. - Abstract: The Monte Carlo method, using the MCNP4C code, was used in this paper to calculate the power distribution in 3-D geometry in the fuel rods of the Syrian Miniature Neutron Source Reactor (MNSR). To normalize the MCNP4C result to the steady state nominal thermal power, the appropriate scaling factor was defined to calculate the power distribution precisely. The maximum power of the individual rod was found in the fuel ring number 2 and was found to be 105 W. The minimum power was found in the fuel ring number 9 and was 79.9 W. The total power in the total fuel rods was 30.9 kW. This result agrees very well with nominal power reported in the reactor safety analysis report which equals 30 kW. Finally, the peak power factors, which are defined as the ratios between the maximum to the average and the maximum to the minimum powers were calculated to be 1.18 and 1.31 respectively.

Deficient Pms2, ERCC1, Ku86, CcOI in field defects during progression to colon cancer.

Science.gov (United States)

Nguyen, Huy; Loustaunau, Cristy; Facista, Alexander; Ramsey, Lois; Hassounah, Nadia; Taylor, Hilary; Krouse, Robert; Payne, Claire M; Tsikitis, V Liana; Goldschmid, Steve; Banerjee, Bhaskar; Perini, Rafael F; Bernstein, Carol

2010-07-28

In carcinogenesis, the "field defect" is recognized clinically because of the high propensity of survivors of certain cancers to develop other malignancies of the same tissue type, often in a nearby location. Such field defects have been indicated in colon cancer. The molecular abnormalities that are responsible for a field defect in the colon should be detectable at high frequency in the histologically normal tissue surrounding a colonic adenocarcinoma or surrounding an adenoma with advanced neoplasia (well on the way to a colon cancer), but at low frequency in the colonic mucosa from patients without colonic neoplasia. Using immunohistochemistry, entire crypts within 10 cm on each side of colonic adenocarcinomas or advanced colonic neoplasias were found to be frequently reduced or absent in expression for two DNA repair proteins, Pms2 and/or ERCC1. Pms2 is a dual role protein, active in DNA mismatch repair as well as needed in apoptosis of cells with excess DNA damage. ERCC1 is active in DNA nucleotide excision repair. The reduced or absent expression of both ERCC1 and Pms2 would create cells with both increased ability to survive (apoptosis resistance) and increased level of mutability. The reduced or absent expression of both ERCC1 and Pms2 is likely an early step in progression to colon cancer. DNA repair gene Ku86 (active in DNA non-homologous end joining) and Cytochrome c Oxidase Subunit I (involved in apoptosis) had each been reported to be decreased in expression in mucosal areas close to colon cancers. However, immunohistochemical evaluation of their levels of expression showed only low to modest frequencies of crypts to be deficient in their expression in a field defect surrounding colon cancer or surrounding advanced colonic neoplasia. We show, here, our method of evaluation of crypts for expression of ERCC1, Pms2, Ku86 and CcOI. We show that frequency of entire crypts deficient for Pms2 and ERCC1 is often as great as 70% to 95% in 20 cm long areas

Anaerobic Digestion of Sugarcane Vinasse Through a Methanogenic UASB Reactor Followed by a Packed Bed Reactor.

Science.gov (United States)

Cabrera-Díaz, A; Pereda-Reyes, I; Oliva-Merencio, D; Lebrero, R; Zaiat, M

2017-12-01

The anaerobic treatment of raw vinasse in a combined system consisting in two methanogenic reactors, up-flow anaerobic sludge blanket (UASB) + anaerobic packed bed reactors (APBR), was evaluated. The organic loading rate (OLR) was varied, and the best condition for the combined system was 12.5 kg COD m -3 day -1 with averages of 0.289 m 3 CH 4  kg COD r -1 for the UASB reactor and 4.4 kg COD m -3 day -1 with 0.207 m 3 CH 4  kg COD r -1 for APBR. The OLR played a major role in the emission of H 2 S conducting to relatively stable quality of biogas emitted from the APBR, with H 2 S concentrations <10 mg L -1 . The importance of the sulphate to COD ratio was demonstrated as a result of the low biogas quality recorded at the lowest ratio. It was possible to develop a proper anaerobic digestion of raw vinasse through the combined system with COD removal efficiency of 86.7% and higher CH 4 and a lower H 2 S content in biogas.

A new safety approach in the design of fast reactors

International Nuclear Information System (INIS)

Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

1987-01-01

A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

Generation IV reactors: international projects

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

2003-01-01

Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

1988-01-01

The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

On alteration of reactor installation (additional installation of No.3 and No.4 plants in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc.)

International Nuclear Information System (INIS)

1985-01-01

The Nuclear Safty Commission sent the reply to the Minister of International Trade and Industry on October 4, 1984, on this matter after having received the report from the Committee on Examination of Nuclear Reactor Safety and carried out the deliberation. It was judged that the applicant has the technical capability required for installing and operating these reactor facilities. Also it was judged that on the safety after these reactor plants are installed, there is no obstacle in the prevention of disaster due to contaminated substances and reactors. The policy of the investigation and deliberation is reported. The contents of the investigation and deliberation are the condition of location such as site, geological features and ground, earthquake, weather, hydraulic problem and social environments, the safety design of reactor facilities, the evaluation of radiation exposure dose in normal operation, the analysis of abnormal transient change in operation, accident analysis and the evaluation of location. (Kako, I.)

Practical course on reactor instrumentation

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-06-01

This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

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

The Simulator Development for RDE Reactor

Science.gov (United States)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.