François de Rose (1910 - 2014)

CERN Multimedia

Corinne Pralavorio

2014-01-01

One of CERN’s founding fathers has passed away.   François de Rose in the ATLAS cavern during his visit to CERN in 2013. Visionaries have the freedom of mind to shape the future when other people’s horizons are obstructed by the present. François de Rose was a visionary. In the aftermath of the Second World War, when Europe was in ruin, when absolutely everything had to be rebuilt, the diplomat understood the importance of reviving fundamental research and, above all, of cooperation on a continental scale as the driving force of this ambition. In a Europe that was just starting to get back on its feet, it would be no mean feat. Nonetheless, François, alongside the prominent physicists of the time, put his energy into making this vision a reality. They lobbied governments for the creation of a centre that would work towards this goal, winning support, and CERN was established in 1954, an achievement of which François was extremely...

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 ...

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.

Outline of additional installation works in Oi Nuclear Power Station

International Nuclear Information System (INIS)

Shibata, Atsuo; Matsuoka, Motokazu

1987-01-01

At present in Oi Nuclear Power Station, Kansai Electric Power Co., Inc., No.1 and No.2 plants of 1175 MWe each are in operation. In order to stabilize power supply for a long term, the additional installation of No.3 and No.4 plants of 1180 MWe each is in progress. The No.3 and No.4 plants are PWRs, for which prestressed concrete reactor containment vessels were adopted, and the start of operation is scheduled in October, 1991 in No.3 plant and in August, 1992 in No.4 plant. In the execution of main construction works, the preservation of the existing plants is the most important. For excavating the tunnels for seawater channels, tunnel boring machines were used, to avoid blasting. Pneumatic caisson method was adopted for a part of the circulating pump building. No vibration, no noise piling method was adopted for respective sheathing construction. The blasting vibration control in the excavation for foundations and the analysis of the behavior of retaining walls are to be carried out by information-oriented work execution. The outline of the main construction works and the preservation of the existing facilities are reported. (Kako, I.)

Structural and electronic properties of low-index stoichiometric BiOI surfaces

Energy Technology Data Exchange (ETDEWEB)

Dai, Wen-Wu; Zhao, Zong-Yan, E-mail: zzy@kmust.edu.cn

2017-06-01

As promising photocatalyst driven by visible-light, BiOI has attracted more and more attention in the past years. However, the surface structure and properties of BiOI that is the most important place for the photocatalytic have not been investigated in details. To this end, density functional theory was performed to calculate the structural and electronic properties of four low-index stoichiometric surfaces of BiOI. It is found that the relaxation of the low-index BiOI surfaces are relatively small, especially the (001) surface. Thus, the surface energies of BiOI are very relatively small. Moreover, there are a few surface states below the bottom of conduction band in the first layer except the (001) surface, which maybe capture the photo-excited carriers. In all of the most stable terminated planes, all the dangling bonds are cleaved from the broken Bi-O bonds. In the case of (001) surface, the dangling bond density of Bi atoms for the (001) surface is zero per square nano. Therefore, the (001) surface is thermodynamically lowest-energy surface of BiOI, and it is the predominant surface (51.4%). As a final remark, the dangling bonds density of bismuth atoms determines not only the surface energy, but also the surface relaxation. Finally, the equilibrium morphology of BiOI was also proposed and provided, which is determined through the Wulff construction. These results will help us to better understand the underlying photocatalytic mechanism that is related to BiOI surfaces, and provide theoretical support for some experimental studies about BiOI-based photocatalyst in future. - Highlights: • Four low-index BiOI surfaces have been calculated by DFT method. • The relaxations of the low-index BiOI surfaces are relatively small. • There are a few surface states below the bottom of conduction band in the first layer. • The dangling bonds density of bismuth atoms determines not only the surface energy, but also the surface relaxation. • The thermodynamic

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.

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

Rapid adsorption properties of flower-like BiOI nanoplates synthesized via a simple EG-assisted solvothermal process

Energy Technology Data Exchange (ETDEWEB)

Zhang, Bin; Ji, Guangbin, E-mail: gbji@nuaa.edu.cn [Nanjing University of Aeronautics and Astronautics, College of Materials Science and Technology (China); Gondal, M. A. [King Fahd University of Petroleum and Minerals, Physics Department (Saudi Arabia); Liu, Yousong; Zhang, Xingmiao; Chang, Xiaofeng; Li, Nianwu [Nanjing University of Aeronautics and Astronautics, College of Materials Science and Technology (China)

2013-07-15

Uniform well-crystallized flower-like BiOI nanoplates contained 3.7 nm mesopores, which may be attributed to the internanosheet spaces of BiOI with maximum pore diameters of about 30 nm, were successfully synthesized via a simple ethylene glycol-assisted solvothermal method. The as-prepared porous BiOI nanoplates exhibited excellent adsorption ability, and the saturated extent of adsorption of BiOI over an RhB solution was as high as 197 mg/g, which is much higher than those for BiOCl and BiOBr prepared via the same method and with a similar surface area. The probable adsorption mechanism could have originated from the interaction between the I atom in BiOI and a proton in RhB at different pH values and temperatures. With visible light irradiation ({lambda} > 420 nm), 80 % of the RhB was degraded in 4 h, while BiOI still demonstrated reasonably outstanding photocatalytic ability under green light ({lambda} = 550 {+-} 15 nm) because of its low-energy gap (1.72 eV). The degradation test for BiOI under irradiation at {lambda} = 550 {+-} 15 nm is an excellent achievement for field applications because the catalyst can be applied in solar irradiation to remove organic pollutants, which may be of great value BiOI complex.

A familial study of Hallermann–Streiff–François syndrome

Directory of Open Access Journals (Sweden)

Epée E

2017-06-01

Full Text Available E Epée,1 D Beleho,2 AT Bitang,3 VA Njami,4 C Bengondo,5 Côme Ebana Mvogo1 1Ophthalmology Department, Yaoundé University Teaching Hospital, Yaoundé, Cameroon; 2Ophthalmology Department, Okola District Hospital, Okola, Cameroon; 3Faculty of Medicine and Biomedical Sciences, University of Yaoundé I, Yaoundé, Cameroon; 4Higher Institute of Health Sciences, Université des Montagnes, Bangangté, Cameroon; 5Stomatology Department, Yaoundé University Teaching Hospital, Yaoundé, Cameroon Abstract: Hallermann–Streiff–François syndrome is a rare sporadic genetic pathology characterized by a phenotype consisting of growth retardation, ocular abnormalities, and a “bird-like head”. We hereby report a case of this syndrome found in three generations of the same family – father, daughter, and grand-daughter – who presented with a short stature and facial dysmorphic features, nystagmus, cataract, and bilateral microphthalmia. The discussion is based on the clinical and genetic aspects, and the challenges in management of this oculo-mandibulo-facial syndrome. The association of congenital cataract, facial dysmorphic features, and microphthalmia, should guide the diagnosis of dysmorphic syndromes such as Hallermann–Streiff–François syndrome. Keywords: Hallermann–Streiff–François syndrome, familial cataract, dysmorphic features, rare, Cameroon

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

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.)

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.

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

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

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 ...

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.

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

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.

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.

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)

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 ...

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

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      

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.

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.)

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, ...

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)

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

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

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.

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

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)

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...

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.

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

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.

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)

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

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.)

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.

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 ...

François Delsarte and Modern Dance: an encounter in physical expression

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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.

Dynamics of TRIGA-3 Salazar Reactor

International Nuclear Information System (INIS)

Gallardo S, L.F.

1990-01-01

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

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.)

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

speech impediments. The results of the research conducted led to the following conclusions: 1. Among pediatric patients with OI there are disorders in the stomatognathic system. The most common dysfunctions are: abdominal, swallowing and sucking disorders, abnormal muscle structure of the rumen and biomechanical disorders in the temporomandibular joints. Breastfeeding significantly contributes to swallowing disorders. 2. The therapeutic process involving children with OI requires the cooperation of specialists in orthopedics, pediatrics, physiotherapy, orthodontics and neurologopedics to carry out comprehensive diagnostics and treatment tailored to the individual needs of the patient. 3. In order to draw final conclusions, there is a need for more research by means of objective tools, such as EMG and a condensate recorder.

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)

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

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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.

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.

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.

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.

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.

Decommissioning of a small reactor (BR3 reactor, Belgium)

International Nuclear Information System (INIS)

Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

2002-01-01

Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (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 ...

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.

Enhanced photocatalytic activity and characterization of magnetic Ag/BiOI/ZnFe2O4 composites for Hg0 removal under fluorescent light irradiation

Science.gov (United States)

Li, Chengwei; Zhang, Anchao; Zhang, Lixiang; Song, Jun; Su, Sheng; Sun, Zhijun; Xiang, Jun

2018-03-01

A series of magnetic Ag/BiOI/ZnFe2O4 hybrids synthesized via hydrothermal process, subsequent deposition-precipitation and photoreduction method were employed to remove elemental mercury (Hg0) under fluorescent light irradiation. The effects of Ag content, fluorescent light irradiation, reaction temperature, pH value, flue gas composition, anions and photocatalyst dosage on Hg0 removal were investigated in detail. The as-synthesized photocatalysts were characterized using N2 adsorption-desorption, XRD, SEM, TEM, HRTEM, XPS, VSM, DRS, ESR, PL and photocurrent response. The results showed that the ternary Ag/BiOI/ZnFe2O4 hybrids possessed enhanced visible-light-responsive photocatalytic performances for Hg0 removal. Ag/BiOI/ZnFe2O4 photocatalyst could be easily recovered from the reaction solution by an extra magnet and was stable in the process of Hg0 removal. Lower content of Ag was highly dispersed on the surface of BiOI/ZnFe2O4, while higher content of Ag would result in some aggregations and/or the blockages of micropore. In comparison to BiOI/ZnFe2O4, Ag deposited BiOI/ZnFe2O4 material showed lower recombination rate of electron-hole pairs. The superior Hg0 oxidation removal could correspond to good match of BiOI and ZnFe2O4, excellent fluidity and surface plasmon resonance effect of Ag0 nanoparticles, which led to higher separation efficiency of photogenerated electrons and holes, thereby enhancing the hybrids' photocatalytic activity.

Integral test of JENDL-3.3 for fast reactors

International Nuclear Information System (INIS)

Chiba, Gou

2003-01-01

An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

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...

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.

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 Dynamic Assessment of the François-Mitterrand Library

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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.

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.

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

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.

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.)

Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

Energy Technology Data Exchange (ETDEWEB)

Gallardo S, L F

1991-12-31

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

Energy Technology Data Exchange (ETDEWEB)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

2016-09-15

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

International Nuclear Information System (INIS)

Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

2013-01-01

The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

Kernbichler, W.; Miley, G.H.; Heindler, M.

1989-01-01

The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

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

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

SILAR BiOI-Sensitized TiO2 Films for Visible-Light Photocatalytic Degradation of Rhodamine B and 4-Chlorophenol.

Science.gov (United States)

Odling, Gylen; Robertson, Neil

2017-04-05

BiOI nanoplates were deposited upon a film of TiO 2 nanoparticles derived from a commercial source using a simple room temperature sequential ionic layer adsorption and reaction (SILAR) method. X-ray diffraction, X-ray photoelectron spectroscopy and electron microscopies have been used to confirm the crystal phase, chemical states of key elements and morphology of the BiOI nanoplate-TiO 2 composites. Using both valence band X-ray photoelectron spectroscopy and UV/Vis diffuse reflectance measurements the band structure of the composites is determined to be that of a type II heterojunction. Through initial screening of the photocatalytic activity of the SILAR-modified films it was determined that five SILAR cycles are optimal in the photocatalytic degradation of rhodamine B. The visible-light sensitisation effect of BiOI was then proven by examination of the photocatalytic degradation of the colourless organic pollutant 4-chlorophenol, showing a large enhancement over an equivalent TiO 2 film. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Removal in a lump of JRR-3 nuclear reactor

International Nuclear Information System (INIS)

Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

1989-01-01

The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

Kakefuda, K.; Tani, M.; Isshiki, M.

1992-01-01

The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

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.

EL3 reactor description and safety analysis report

International Nuclear Information System (INIS)

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10 14 neutrons/cm 2 /sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements [fr

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2017-07-15

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

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.

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.

3 Investment Scenarios for Fast Reactors

International Nuclear Information System (INIS)

Shoai Tehrani, Bianka; Da Costa, Pascal

2013-01-01

Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

D-3He fueled FRC reactor 'ARTEMIS-L'

International Nuclear Information System (INIS)

Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

1992-09-01

A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

Description of the RA-3 research reactor as a model facility

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

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 simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

1999-01-01

Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

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,...

Integral test of JENDL-3.3 on fast reactors

International Nuclear Information System (INIS)

Chiba, Gou; Hazama, Taira

2003-05-01

An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

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...

3D CAD model of the subcritical nuclear reactor of IPN

International Nuclear Information System (INIS)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

2016-09-01

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (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

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

Estimation of reactor pool water temperature after shutdown in JRR-3M

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

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.

Modelling of MOCVD Reactor: New 3D Approach

Science.gov (United States)

Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

2014-04-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Modelling of MOCVD reactor: new 3D approach

International Nuclear Information System (INIS)

Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

2014-01-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

Science.gov (United States)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

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…

Operating reactors licensing actions summary. Vol. 3, No. 3

International Nuclear Information System (INIS)

1983-04-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory 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 Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

Report on generation IV technical working group 3 : liquid metal reactors

International Nuclear Information System (INIS)

Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

2002-01-01

This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

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

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

EL-3 dismantling of an experimental reactor

International Nuclear Information System (INIS)

1989-01-01

The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

1992-01-01

The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

Characteristics of D-3He fueled frc reactor: ARTEMIS-L

International Nuclear Information System (INIS)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

Summary of the 3rd workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Summary of the 3rd workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

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.

Osteogenesis imperfecta type 3 in South Africa: Causative mutations in FKBP10

Directory of Open Access Journals (Sweden)

Alvera Vorster

2017-05-01

Full Text Available Background. A relatively high frequency of autosomal recessively inherited osteogenesis imperfecta (OI type 3 (OI-3 is present in the indigenous black southern African population. Affected persons may be severely handicapped as a result of frequent fractures, progressive deformity of the tubular bones and spinal malalignment. Objective. To delineate the molecular basis for the condition. Methods. Molecular investigations were performed on 91 affected persons from seven diverse ethnolinguistic groups in this population. Results. Following polymerase chain reaction amplification and direct cycle sequencing, FKBP10 mutations were identified in 45.1% (41/91 OI-3-affected persons. The homozygous FKBP10 c.831dupC frameshift mutation was confirmed in 35 affected individuals in the study cohort. Haplotype analysis suggests that this mutation is identical among these OI-3-affected persons by descent, thereby confirming that they had a common ancestor. Compound heterozygosity of this founder mutation was observed, in combination with three different deleterious FKBP10 mutations, in six additional persons in the cohort. Four of these individuals had the c.831delC mutation. Conclusion. The burden of the disorder, both in frequency and severity, warrants the establishment of a dedicated service for molecular diagnostic confirmation and genetic management of persons and families with OI in southern Africa.

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

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

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

CERN Document Server

Chiba, G

2002-01-01

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

Development of 3D CFD simulation method in nuclear reactor safety analysis

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

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

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

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.

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

Energy Technology Data Exchange (ETDEWEB)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

International Nuclear Information System (INIS)

Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

1982-03-01

A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Systems analysis of the CANDU 3 Reactor

Energy Technology Data Exchange (ETDEWEB)

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

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

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.)

Ignition access in a D-3He helical reactor

International Nuclear Information System (INIS)

Mitarai, Osamu

2003-01-01

Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

Fusion blankets for catalyzed D--D and D--He3 reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

Fusion blankets for catalyzed D--D and D--3He reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Prolyl 3-hydroxylase 1 and CRTAP are mutually stabilizing in the endoplasmic reticulum collagen prolyl 3-hydroxylation complex.

Science.gov (United States)

Chang, Weizhong; Barnes, Aileen M; Cabral, Wayne A; Bodurtha, Joann N; Marini, Joan C

2010-01-15

Null mutations in cartilage-associated protein (CRTAP) and prolyl 3-hydroxylase 1 (P3H1/LEPRE1) cause types VII and VIII OI, respectively, two novel recessive forms of osteogenesis imperfecta (OI) with severe to lethal bone dysplasia and overmodification of the type I collagen helical region. CRTAP and P3H1 form a complex with cyclophilin B (CyPB) in the endoplasmic reticulum (ER) which 3-hydroxylates the Pro986 residue of alpha1(I) and alpha1(II) collagen chains. We investigated the interaction of complex components in fibroblasts from types VII and VIII OI patients. Both CRTAP and P3H1 are absent or reduced on western blots and by immunofluorescence microscopy in cells containing null mutations in either gene. Levels of LEPRE1 or CRTAP transcripts, however, are normal in CRTAP- or LEPRE1-null cells, respectively. Stable transfection of a CRTAP or LEPRE1 expression construct into cells with null mutations for the transfected cDNA restored both CRTAP and P3H1 protein levels. Normalization of collagen helical modification in transfected CRTAP-null cells demonstrated that the restored proteins functioned effectively as a complex. These data indicate that CRTAP and P3H1 are mutually stabilized in the collagen prolyl 3-hydroxylation complex. CyPB levels were unaffected by mutations in either CRTAP or LEPRE1. Proteasomal inhibitors partially rescue P3H1 protein in CRTAP-null cells. In LEPRE1-null cells, secretion of CRTAP is increased compared with control cells and accounts for 15-20% of the decreased CRTAP detected in cells. Thus, mutual stabilization of P3H1 and CRTAP in the ER collagen modification complex is an underlying mechanism for the overlapping phenotype of types VII and VIII OI.

GRIMH3: A new reactor calculation code at Savannah River Site

International Nuclear Information System (INIS)

Le, T.T.; Pevey, R.E.

1993-01-01

The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Sequential Oxygenation Index and Organ Dysfunction Assessment within the First 3 Days of Mechanical Ventilation Predict the Outcome of Adult Patients with Severe Acute Respiratory Failure

Directory of Open Access Journals (Sweden)

Hsu-Ching Kao

2013-01-01

Full Text Available Objective. To determine early predictors of outcomes of adult patients with severe acute respiratory failure. Method. 100 consecutive adult patients with severe acute respiratory failure were evaluated in this retrospective study. Data including comorbidities, Sequential Organ Failure Assessment (SOFA score, Acute Physiological Assessment and Chronic Health Evaluation II (APACHE II score, PaO2, FiO2, PaO2/FiO2, PEEP, mean airway pressure (mPaw, and oxygenation index (OI on the 1st and the 3rd day of mechanical ventilation, and change in OI within 3 days were recorded. Primary outcome was hospital mortality; secondary outcome measure was ventilator weaning failure. Results. 38 out of 100 (38% patients died within the study period. 48 patients (48% failed to wean from ventilator. Multivariate analysis showed day 3 OI ( and SOFA ( score were independent predictors of hospital mortality. Preexisting cerebrovascular accident (CVA ( was the predictor of weaning failure. Results from Kaplan-Meier method demonstrated that higher day 3 OI was associated with shorter survival time (log-Rank test, . Conclusion. Early OI (within 3 days and SOFA score were predictors of mortality in severe acute respiratory failure. In the future, prospective studies measuring serial OIs in a larger scale of study cohort is required to further consolidate our findings.

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.

‘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.

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)

Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

International Nuclear Information System (INIS)

Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

2012-01-01

10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

Safety in the ARIES-III D-3He tokamak reactor design

International Nuclear Information System (INIS)

Herring, J.S.; Dolan, T.J.

1992-01-01

This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

Reversed field pinch reactor study 3

International Nuclear Information System (INIS)

Hollis, A.A.; Mitchell, J.T.D.

1977-12-01

This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

2003-01-01

The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

The World's Reactors no. 70 - Forsmark 3, BWR-75

International Nuclear Information System (INIS)

Anon.

1976-01-01

A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

[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'.

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

International Nuclear Information System (INIS)

1979-11-01

This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

Development of telerobotic systems for reactor decommissioning, (3)

International Nuclear Information System (INIS)

Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

1991-01-01

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

2007-11-01

The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

International Nuclear Information System (INIS)

El-Sahlamy, N.M.

2017-01-01

One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

The qualification of U3O8 as research reactor fuel

International Nuclear Information System (INIS)

Krull, W.

1983-01-01

This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

2004-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

International Nuclear Information System (INIS)

Khan, Jahirul Haque

2013-01-01

The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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...

3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

International Nuclear Information System (INIS)

Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

1974-01-01

1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

A robot-automated work site for repair of the Chinon A3 reactor

International Nuclear Information System (INIS)

Raynal, A.

1987-01-01

In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time [fr

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.

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

Investigations related to a one-piece removal of the reactor block in the frame of the JRR-3 reconstruction program

International Nuclear Information System (INIS)

Onishi, N.; Kanenari, A.; Futamura, Y.; Sakurai, H.; Suzuki, S.; Nagase, T.; Iwatani, A.; Otsubo, F.

1987-01-01

In the Japan Atomic Energy Research Institute (JAERI), an outdated research reactor (Japan Research Reactor No.3; JRR-3) was removed to a storage facility between October 14th and November 7th, 1986. The removal of the 2250-ton reactor block (10 x 10 x 10 m) was performed as a part of a program to replace the JRR-3's core (10-MW thermal) with an upgraded research reactor core. The heavy water and fuel elements were taken out from the JRR-3 before removal work began. The reactor block was raised about 3.7 meters, using a 12-cubic meter steel frame and a center-hole jack system. The reactor block was then transported horizontally about 34 meters on steel rails, using four 100-ton jacks, to a storage facility. Finally, the reactor block was lowered 14 meters into the storage facility. After the reactor block was stored, a new 20-MW thermal, light-water moderated and cooled JRR-3 core will be built, with criticality targeted for 1989

Decommissioning of the BR3 pressurized-water reactor

International Nuclear Information System (INIS)

Massaut, V.

1996-01-01

The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

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

International Nuclear Information System (INIS)

1966-01-01

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

Recent developments in osteogenesis imperfecta [version 1; referees: 3 approved

Directory of Open Access Journals (Sweden)

Joseph L. Shaker

2015-09-01

Full Text Available Osteogenesis imperfecta (OI is an uncommon genetic bone disease associated with brittle bones and fractures in children and adults. Although OI is most commonly associated with mutations of the genes for type I collagen, many other genes (some associated with type I collagen processing have now been identified. The genetics of OI and advances in our understanding of the biomechanical properties of OI bone are reviewed in this article. Treatment includes physiotherapy, fall prevention, and sometimes orthopedic procedures. In this brief review, we will also discuss current understanding of pharmacologic therapies for treatment of OI.

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.)

Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Advances in Reactor physics, mathematics and computation. Volume 3

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.

1994-01-01

TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

On the major DYN3D developments for fast reactor design and transient analysis

International Nuclear Information System (INIS)

Merk, B.; Kliem, S.

2013-01-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

On the major DYN3D developments for fast reactor design and transient analysis

Energy Technology Data Exchange (ETDEWEB)

Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

2013-07-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

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.

RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report.

Operating reactors licensing actions summary. Vol. 3, No. 6

International Nuclear Information System (INIS)

1983-07-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Study of seismic responses of Candu-3 reactor building using isolator bearings

International Nuclear Information System (INIS)

Biswas, J.K.

1992-01-01

Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

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

1988-01-01

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

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

Research on intelligent monitor for 3D power distribution of reactor core

International Nuclear Information System (INIS)

Xia, Hong; Li, Bin; Liu, Jianxin

2014-01-01

Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

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.

Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

International Nuclear Information System (INIS)

Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

2017-01-01

Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

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

Energy Technology Data Exchange (ETDEWEB)

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

1958-07-01

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

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. 

3. International conference on catalysis in membrane reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

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)

Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

International Nuclear Information System (INIS)

Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

2003-01-01

Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

Technical report on implementation of reactor internal 3D modeling and visual database system

International Nuclear Information System (INIS)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun

1996-06-01

In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation's integrated computer aided engineering system, such as Mitsubishi's NUWINGS (Japan), AECL's CANDID (Canada) and Duke Power's PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new

Technical report on implementation of reactor internal 3D modeling and visual database system

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

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

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Development of a 3-D flow analysis computer program for integral reactor

International Nuclear Information System (INIS)

Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

2003-01-01

A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

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

Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

International Nuclear Information System (INIS)

Chi-Mei Lee; Kuo-Tsang Hung

2006-01-01

The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

International Nuclear Information System (INIS)

1992-04-01

Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

1992-01-01

The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

Annealing of the BR3 reactor pressure vessel

International Nuclear Information System (INIS)

Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

1985-01-01

The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

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

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

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

Master-3.0: multi-purpose analyzer for static and transient effects of reactors

International Nuclear Information System (INIS)

Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

2002-03-01

MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

3 D flow computations under a reactor vessel closure head

International Nuclear Information System (INIS)

Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

1995-12-01

The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

2012-03-15

JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (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.

Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (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.

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

3D CFD for chemical transport profiles in a rotating disk CVD reactor

Science.gov (United States)

Han, Jong-Hyun; Yoon, Do-Young

2010-06-01

The RDCVD (Rotating Disk Chemical Vapor Deposition) technique is an appropriate method for uniform deposition of grains, such as compound semiconductior materials. The substrate temperature and rotation speed are the major factors, which determine the thickness uniformity of the deposited films. This paper investigates 3D CFD (3 Dimensional Computational Fluid Dynamics) simulation results of flow and heat transfer in a reactor of RDCVD using Fluent. In order to establish the reducibility of buoyancy effect on deposition quality, the chemical transport profile upon the disk heated is examined successfully in 3D domain for different rotating speeds. The resulting vortex flows due the simultaneous buoyance and centrifuge are discussed qualitatively in the 3D virtual system of a RDCVD reactor. 3D CFD is even more effective to describe the internal vortex flows due to the competitive inlet, buoyancy and centrifuge flows, which cannot be realized in the general 2D (2 Dimensional) CFD.[Figure not available: see fulltext.

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.

Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

2012-07-15

Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

KAUST Repository

Karsenty, Florent

2012-08-16

Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

International Nuclear Information System (INIS)

Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

2015-01-01

Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

Aerosol synthesis and characterization of nanostructured particles of Y3Al5O12:Ce3+ and Y2O3:Eu3+

Directory of Open Access Journals (Sweden)

Marinković Katarina R.

2007-01-01

Full Text Available Nanostructured YAG:Ce3+ and Y2O3:Eu3+ were synthesized by low temperature (320°C aerosol synthesis-LTAS and high temperature (900°C aerosol synthesis-HTAS, respectively. The synthesis included aerosol generation from a nitrate precursor solution by an ultrasonic atomizer (1.3 MHz. The obtained aerosol was introduced into a tubular flow reactor, using air as the carrier gas, where successively, on a droplet level, evaporation/drying, precipitation and thermolysis occurred. The obtained powders were collected and thermally treated at different temperatures (900-1200°C. The phase development and the morphology were investigated by the X-ray powder diffraction method (XRPD and scanning electron microscopy combined with energy dispersive spectrometry (SEM/EDS. Structural refinement was performed using the Rietveld method with the Fullprof and Koalariet programs. The average crystallite size for the Y2O3:Eu system was calculated using the Profit program. It was shown that 89 wt.% of Y3Ai5Oi2:Ce was obtained by annealing (1000°C/6 h the as prepared, amorphous powder, synthesized by the low temperature aerosol method (LTAS. High temperature spray pyrolysis (HTAS at 900°C led to the formation of the targeted cubic phase of Y2O3:Eu3+. The microstructural parameters of the asprepared samples of the Y2O3:Eu3+ system indicate the formation of nanostructures with crystallite size smallest than 20 nm. The substitution of luminescent centers (Ce3+, Eu3+ into a host lattice (YAG, Y2O3, respectively was confirmed by changes in the crystal lattice parameters. Also, it was shown in both systems that good morphological characteristics (non-a­gglomerated, spherical, submicron particles were obtained enabling improved luminescent characteristics.

Study on effects of development of reactor constant in fast reactor analysis

International Nuclear Information System (INIS)

Chiba, Gou

2002-12-01

Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

Energy Technology Data Exchange (ETDEWEB)

Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

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

International Nuclear Information System (INIS)

2015-01-01

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

Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

International Nuclear Information System (INIS)

Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

2017-01-01

Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

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

Tritium production, management and its impact on safety for a D-3He fusion reactor

International Nuclear Information System (INIS)

Sze, D.K.; Herring, S.; Sawan, M.

1991-11-01

About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

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

Characteristics of UV-MicroO3 Reactor and Its Application to Microcystins Degradation during Surface Water Treatment

Directory of Open Access Journals (Sweden)

Guangcan Zhu

2015-01-01

Full Text Available The UV-ozone (UV-O3 process is not widely applied in wastewater and potable water treatment partly for the relatively high cost since complicated UV radiation and ozone generating systems are utilized. The UV-microozone (UV-microO3, a new advanced process that can solve the abovementioned problems, was introduced in this study. The effects of air flux, air pressure, and air humidity on generation and concentration of O3 in UV-microO3 reactor were investigated. The utilization of this UV-microO3 reactor in microcystins (MCs degradation was also carried out. Experimental results indicated that the optimum air flux in the reactor equipped with 37 mm diameter quartz tube was determined to be 18∼25 L/h for efficient O3 generation. The air pressure and humidity in UV-microO3 reactor should be low enough in order to get optimum O3 output. Moreover, microcystin-RR, YR, and LR (MC-RR, MC-YR, and MC-LR could be degraded effectively by UV-microO3 process. The degradation of different MCs was characterized by first-order reaction kinetics. The pseudofirst-order kinetic constants for MC-RR, MC-YR, and MC-LR degradation were 0.0093, 0.0215, and 0.0286 min−1, respectively. Glucose had no influence on MC degradation through UV-microO3. The UV-microO3 process is hence recommended as a suitable advanced treatment method for dissolved MCs degradation.

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

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

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

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

International Nuclear Information System (INIS)

Reyes F, M. C.; Del Valle G, E.; Filio L, C.

2013-10-01

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

2008-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

International Nuclear Information System (INIS)

Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

2013-01-01

Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

International Nuclear Information System (INIS)

Greene, S.R.; Fisher, S.E.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

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)

A 3D transport-based core analysis code for research reactors with unstructured geometry

International Nuclear Information System (INIS)

Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

2013-01-01

Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

A case study of aerosol data assimilation with the Community Multi-scale Air Quality Model over the contiguous United States using 3D-Var and optimal interpolation methods

Directory of Open Access Journals (Sweden)

Y. Tang

2017-12-01

Full Text Available This study applies the Gridpoint Statistical Interpolation (GSI 3D-Var assimilation tool originally developed by the National Centers for Environmental Prediction (NCEP, to improve surface PM2.5 predictions over the contiguous United States (CONUS by assimilating aerosol optical depth (AOD and surface PM2.5 in version 5.1 of the Community Multi-scale Air Quality (CMAQ modeling system. An optimal interpolation (OI method implemented earlier (Tang et al., 2015 for the CMAQ modeling system is also tested for the same period (July 2011 over the same CONUS. Both GSI and OI methods assimilate surface PM2.5 observations at 00:00, 06:00, 12:00 and 18:00 UTC, and MODIS AOD at 18:00 UTC. The assimilations of observations using both GSI and OI generally help reduce the prediction biases and improve correlation between model predictions and observations. In the GSI experiments, assimilation of surface PM2.5 (particle matter with diameter < 2.5 µm leads to stronger increments in surface PM2.5 compared to its MODIS AOD assimilation at the 550 nm wavelength. In contrast, we find a stronger OI impact of the MODIS AOD on surface aerosols at 18:00 UTC compared to the surface PM2.5 OI method. GSI produces smoother result and yields overall better correlation coefficient and root mean squared error (RMSE. It should be noted that the 3D-Var and OI methods used here have several big differences besides the data assimilation schemes. For instance, the OI uses relatively big model uncertainties, which helps yield smaller mean biases, but sometimes causes the RMSE to increase. We also examine and discuss the sensitivity of the assimilation experiments' results to the AOD forward operators.

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Plutonium-burn high temperature gas-cooled reactor for 3E+3S

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst

Directory of Open Access Journals (Sweden)

New Pei Yee

2008-04-01

Full Text Available A one-dimensional mathematical model was developed to simulate the performance of catalytic fixed bedreactor for carbon dioxide reforming of methane over Rh/Al2O3 catalyst at atmospheric pressure. The reactionsinvolved in the system are carbon dioxide reforming of methane (CORM and reverse water gas shiftreaction (RWGS. The profiles of CH4 and CO2 conversions, CO and H2 yields, molar flow rate and molefraction of all species as well as reactor temperature along the axial bed of catalyst were simulated. In addition,the effects of different reactor temperature on the reactor performance were also studied. The modelscan also be applied to analyze the performances of lab-scale micro reactor as well as pilot-plant scale reactorwith certain modifications and model verification with experimental data. © 2008 BCREC UNDIP. All rights reserved.[Received: 20 August 2008; Accepted: 25 September 2008][How to Cite: N.A.S. Amin, I. Istadi, N.P. Yee. (2008. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst. Bulletin of Chemical Reaction Engineering and Catalysis, 3 (1-3: 21-29. doi:10.9767/bcrec.3.1-3.19.21-29

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.)

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

RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

International Nuclear Information System (INIS)

Zecevic, V.

1963-12-01

This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

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

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

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

Reactor safety study applied to the Forsmark 3 Power Plant

International Nuclear Information System (INIS)

Ericsson, G.; Tiren, L.I.

1978-01-01

A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

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

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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.

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

Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E

1985-07-01

A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

1997-01-01

The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

International Nuclear Information System (INIS)

Kishimoto, Katsumi; Arigane, Kenji

2005-03-01

Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18 (case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes. Total activation activity of reactor by the revaluation depended on control rods, thermal shield plates and horizontal experimental tubes, and the value after 1 year from the permanent shutdown of reactor was 1.9x10 14 Bq. (author)

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

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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.

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

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

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

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

Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

International Nuclear Information System (INIS)

Uegata, T.; Saji, E.; Tanaka, H.

1993-01-01

Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

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)

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)

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.)

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

Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Energy Technology Data Exchange (ETDEWEB)

Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

1985-07-01

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

Energy Technology Data Exchange (ETDEWEB)

C. B. Davis

2006-07-01

The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

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)

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

Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

2004-01-01

The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

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

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

International Nuclear Information System (INIS)

Hernandez, N.; Alonso, G.; Valle, E. del

2004-01-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

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

Survey of group data libraries for use of the DYN3D program for WWER type reactors

International Nuclear Information System (INIS)

Mittag, S.

1994-06-01

So-called few-group neutron data have to be used as input data in core models (such as DYN3D) calculating the reactor behaviour. A survey is given of qualified data libraries for the reactor cores of Russian VVER. The information about primary data used in group data generation and the accuracy reached by the cell codes is compiled in tables. To assess the quality of the data, comparisons have been made between measured and calculated reactor parameters. The information available does not show significant differences concerning the quality of the data libraries. (orig.) [de

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.

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)

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

Validation and application of 3D-methods for the design and safety analysis of high temperature reactors

International Nuclear Information System (INIS)

Bader, J.; Lapins, J.; Buck, M; Bernnat, W.; Laurien, E.

2011-01-01

Some of the concepts for future nuclear reactors are high-temperature gas-cooled reactors. Previous simulation codes for their cores were often based on one- or two-dimensional models, but today's increasing computer capabilities make an advance to 3D-codes possible now. Our thermal-hydraulic code ATTICA3D (Advanced Thermal-hydraulic Tool for In-vessel and Core Analysis in 3 Dimensions) is based on the porous media approach, including 3-D models of heat conduction and gas flow, using a coarse-grid integration method for the time-dependent conservation equations of mass, momentum and energy. Results of numerical calculations for various validation cases are presented: First, the test facility SANA is chosen, which has been used to study heat transfer phenomena inside a coolant-gas filled pebble-bed core, which was heated by embedded electrical heating elements. Calculations were carried out for different tests taken from the experimental database. Measured and calculated temperatures at different positions are compared and found in good agreement. Second, our code was used to simulate a depressurized loss of forced cooling experiment with simulated decay heat in the AVR Experimental Reactor. Due to its design with the shut-down rods located inside columnar noses, which extend into the pebble bed of the core, geometry and power distribution are genuinely three-dimensional. The power distribution was calculated by the 3D-Neutronic Diffusion Code CITATION in conjunction with the spectral code MICROX-2. The neutronics and thermal-hydraulics calculations were carried out for a 3D, 45°-degree section of the reactor. It is demonstrated, that the experimental results could be qualitatively reproduced. (author)

VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

International Nuclear Information System (INIS)

Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

1983-05-01

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-01

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

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...

Atomization of U3Si2 for research reactor fuel

International Nuclear Information System (INIS)

Kim, C.K.; Kim, K.H.; Lee, C.T.; Kuk, I.H.

1995-01-01

Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U 3 Si and the mixture of U 3 Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U 3 Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

Science.gov (United States)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

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

2012-01-01

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

DRAGON 3.05D, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: The computer code DRAGON contains a collection of models that can simulate the neutron behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. 2 - Methods: The code DRAGON contains a multigroup flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries. The execution of DRAGON is

Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

1998-03-01

The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

Detailed modeling of KALININ-3 NPP VVER-1000 reactor pressure vessel by the coupled system code ATHLET/BIPR-VVER

International Nuclear Information System (INIS)

Nikonov, S.P.; Velkov, K.; Pautz, A.

2011-01-01

The paper gives an overview of the recent developments of a new reactor pressure vessel (RPV) model of VVER-1000 for the coupled system code ATHLET/BIPR-VVER. Based on the previous experience a methodology is worked out for modeling the RPV in a pseudo-3D way with the help of a multiple parallel thermal-hydraulic channel scheme that follows the hexagonal fuel assembly structure from the bottom to the top of the reactor. The results of the first application of the new modeling are discussed on the base of the OECD/NEA coupled code benchmark for Kalinin-3 NPP transient. Coolant mass flow distributions in reactor volume of VVER 1000 reactor are presented and discussed. It is shown that along the core height a mass flow re-distribution of the coolant takes place starting approximately at an axial layer located 1 meter below the core outlet. (author)

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)

Comparative studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 data libraries on the Monte Carlo continuous energy modeling of the gas turbine-modular helium reactor operating with thorium fuels

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile 232 Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of 239 Pu, 233 U and 235 U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B. (author)

3 MW TRIGA Research Reactor facility of BAEC and its Utilization

International Nuclear Information System (INIS)

Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z.

1999-01-01

The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

Verification of RBMK-1500 reactor main circulation circuit model with Cathare V1.3L

International Nuclear Information System (INIS)

Jasiulevicius, A.

2001-01-01

Among other computer codes, French code CATHARE is also applied for RBMK reactor calculations. In this paper results of such application for Ignalina NPP reactor (RBMK-1500 type) main circulation circuit are presented. Three transients calculations were performed: all main circulation pumps (MCP) trip, trip of one main circulation pump and trip of one main circulation pump without a closure of check valve on the pump line. Calculation results were compared to data from the Ignalina NPP, where all these transients were recorded in the years 1986, 1996 and 1998. The presented studies prove the capability of the CATHARE code to treat thermal-hydraulic transients with a reactor scram in the RBMK, in case of single or multiple pump trips. However, the presented model needs further improvements in order to simulate loss of coolant accidents. For this reason, emergency core cooling system should be included in the model. Additional model improvement is also needed in order to gain more independent pressure behavior in both loops. Also, flow rates through the reactor channels should be modeled by dividing channels into several groups, referring to channel power (in RBMK power produced in a channel, located in different parts of the core is not the same). The point-neutron kinetic model of the CATHARE code is not suitable to predict transients when the reactor is operating at a nominal power level. Such transients would require the use of 3D-neutron kinetics model to describe properly the strong space-time effect on the power distribution in the reactor core

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

International Nuclear Information System (INIS)

Till, C.E.

1990-01-01

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

Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

International Nuclear Information System (INIS)

Lerner, Ana Maria; Madariaga, Marcelo

2002-01-01

The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

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

Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

International Nuclear Information System (INIS)

Toft, P.; Borring, J.; Adolph, E.

1988-01-01

A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

Processing requirements for property optimization of Eu2O3-W cermets for fast reactor neutron absorber applications

International Nuclear Information System (INIS)

Pasto, A.E.; Tennery, V.J.

1977-01-01

Europium sesquioxide is a candidate fast reactor neutron absorber material. It possesses several desirable characteristics for this application, but has a low thermal conductivity. This gives rise to pellet cracking during reactor operation. To increase the thermal conductivity without great sacrifice in nuclear worth, addition of tungsten to Eu 2 O 3 has been evaluated. Synthesis and fabrication techniques described allow preparation of high density compacts of Eu 2 O 3 -15 vol. percent tungsten, possessing favorable thermal conductivity and thermal expansion characteristics

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.)

Integrin Beta 3 Regulates Cellular Senescence by Activating the TGF-β Pathway

Directory of Open Access Journals (Sweden)

Valentina Rapisarda

2017-03-01

Full Text Available Cellular senescence is an important in vivo mechanism that prevents the propagation of damaged cells. However, the precise mechanisms regulating senescence are not well characterized. Here, we find that ITGB3 (integrin beta 3 or β3 is regulated by the Polycomb protein CBX7. β3 expression accelerates the onset of senescence in human primary fibroblasts by activating the transforming growth factor β (TGF-β pathway in a cell-autonomous and non-cell-autonomous manner. β3 levels are dynamically increased during oncogene-induced senescence (OIS through CBX7 Polycomb regulation, and downregulation of β3 levels overrides OIS and therapy-induced senescence (TIS, independently of its ligand-binding activity. Moreover, cilengitide, an αvβ3 antagonist, has the ability to block the senescence-associated secretory phenotype (SASP without affecting proliferation. Finally, we show an increase in β3 levels in a subset of tissues during aging. Altogether, our data show that integrin β3 subunit is a marker and regulator of senescence.

TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

Science.gov (United States)

Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

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)

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.)

Evaluation of the trial design studies for an advanced marine reactor, (3)

International Nuclear Information System (INIS)

Ambo, Noriaki; Yokomura, Takeyoshi.

1988-03-01

JAERI have carried out four core designs for three different type reactors (Semi-Integrated, Integrated and Integrated (self-pressured) type reactors), as the trial designs of an Advanced Marine Reactor for three years (1983 ∼ 1985). This report describes the result of comparison and studies of the core specific characteristics of these four cores, which include core concept, specifications, core life, specific power density, burn-up, reactivity control and etc. In conclusion, it was found that the Integrated type reactor core and the Semi-Integrated type reactor core designs satisfy the conditions of long core life (four years), high specific power density (50 ∼ 61 kw/l) and high burn-up (30,000 ∼ 32,000 MWD/t), so these two cores will be optimum designs based on the present technologies. (author)

Development of multi-physics code systems based on the reactor dynamics code DYN3D

Energy Technology Data Exchange (ETDEWEB)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

2011-07-15

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Development of multi-physics code systems based on the reactor dynamics code DYN3D

International Nuclear Information System (INIS)

Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

2011-01-01

The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

International Nuclear Information System (INIS)

Jeong, Jae Jun; Chung, Bub Dong

2005-09-01

For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

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)

Power measurement of the RA-3 reactor using the neutron noise technique and {sup 16}N; Medicion de la potencia del reactor RA-3, mediante la tecnica de ruido neutronico y nitrogeno 16

Energy Technology Data Exchange (ETDEWEB)

Gomez, Angel [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. Reactores y Centrales Nucleares

2003-07-01

This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the {sup 16}N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the {sup 16}N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

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

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

Energy Technology Data Exchange (ETDEWEB)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

2013-07-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

International Nuclear Information System (INIS)

Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

2013-01-01

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

International Nuclear Information System (INIS)

Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

2015-01-01

Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

Classification of micro-CT images using 3D characterization of bone canal patterns in human osteogenesis imperfecta

Science.gov (United States)

Abidin, Anas Z.; Jameson, John; Molthen, Robert; Wismüller, Axel

2017-03-01

Few studies have analyzed the microstructural properties of bone in cases of Osteogenenis Imperfecta (OI), or `brittle bone disease'. Current approaches mainly focus on bone mineral density measurements as an indirect indicator of bone strength and quality. It has been shown that bone strength would depend not only on composition but also structural organization. This study aims to characterize 3D structure of the cortical bone in high-resolution micro CT images. A total of 40 bone fragments from 28 subjects (13 with OI and 15 healthy controls) were imaged using micro tomography using a synchrotron light source (SRµCT). Minkowski functionals - volume, surface, curvature, and Euler characteristics - describing the topological organization of the bone were computed from the images. The features were used in a machine learning task to classify between healthy and OI bone. The best classification performance (mean AUC - 0.96) was achieved with a combined 4-dimensional feature of all Minkowski functionals. Individually, the best feature performance was seen using curvature (mean AUC - 0.85), which characterizes the edges within a binary object. These results show that quantitative analysis of cortical bone microstructure, in a computer-aided diagnostics framework, can be used to distinguish between healthy and OI bone with high accuracy.

A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

International Nuclear Information System (INIS)

Guigon, B.; Vacelet, H.; Dornbusch, D.

2000-01-01

Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

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

Future fuel cycle and reactor strategies. Key issue paper no. 3

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The scope of this paper includes those issues that are expected to remain or become important in the time period from 2015 to 2050. Events in this time frame are difficult to predict with any certainty, so the framework of this paper is necessarily somewhat speculative. The paper includes consideration of all facets of nuclear energy utilization, from ore extraction to final disposal of waste products. The paper first addresses the factors influencing the choice of reactor and fuel cycle. It then goes on to address the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel in various forms. The fast reactor type is then discussed both as stand-alone technology and as technology used in combination with thermal reactors. Thorium fuel use is discussed briefly. This paper is concentrated on the ``medium variant`` energy growth scenario identified in Key Issue Paper No. 1. The effects of either higher or lower growth could, of course, profoundly change the future development of the nuclear power industry. 31 refs, 5 figs, 4 tabs.

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)

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.)

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

International Nuclear Information System (INIS)

Chvetsov, I.; Volkov, A.

2000-01-01

For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

A packed bed membrane reactor for the oxidative dehydrogenation of propane on a Ga2O3 / MoO3 based catalyst

NARCIS (Netherlands)

Kotanjac, Ž.S.; Sint Annaland, van M.; Kuipers, J.A.M.

2010-01-01

Oxidative dehydrogenation of propane has been studied over a Ga2O3/MoO3 based catalyst. Using a differentially operated packed bed reactor with premixed oxygen and propane feed, the kinetic parameters for the main reaction and the consecutive and parallel reactions were experimentally determined. It

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

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

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

International Nuclear Information System (INIS)

Joo, Young Sang; Lee, Jae Han

2007-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Lee, Jae Han

2007-02-15

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

Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

International Nuclear Information System (INIS)

Viais J, J.

1994-01-01

Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

CFX-10 and RELAP5-3D simulations of coolant mixing phenomena in RPV of VVER-1000 reactors

International Nuclear Information System (INIS)

Terzuoli, F.; Moretti, F.; Melideo, D.; D'Auria, F.; Shkarupa, O.

2006-01-01

The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed with the ANSYS CFX-10 CFD code and with the RELAP5-3D system code. In particular, the attention focused on the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. The results have been compared against experimental data from V1000CT-2 Benchmark. (author)

Photocatalytic reactors for treating water pollution with solar illumination, Part 3: a simplified analysis for recirculating reactors

Energy Technology Data Exchange (ETDEWEB)

Sagawe, G.; Bahnemann, D. [Hannover Univ. (Germany). Inst. fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [Universidad Nacional de Litoral, Santa Fe (Argentina). Inst. de Desarrollo Tecnologico para la Imdustria Quimica

2004-11-01

A solar photoreactor operated in the batch, recirculating mode is analyzed in terms of very simple observable variables such as the impinging photon flux, the incident area, the initial concentration, the flow rate, the reactor volume and a property defined as the Observed Photonic Efficiency. The proposed equipment is made of a tubular reactor, a tank, a pump and the connecting pipes. The analysis is formulated in terms of the photon input corresponding to an equivalent batch system that is derived as a new reaction coordinate for photoreactions. Employing several plausible approximations, the pollutant concentration evolution in the tank is cast in terms of very simple analytical solutions. Process photonic efficiencies are defined for the system operation and calculated with respect to the maximum achievable yield corresponding to the differential operation of the solar recirculating reactor. (Author)

Irradiation of an uranium silicide prototype in RA-3 reactor

International Nuclear Information System (INIS)

Calabrese, R.; Estrik, G.; Notari, C.

1996-01-01

The factibility of irradiation of an uranium silicide (U 3 Si 2 ) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U 3 O 8 as fissible material. The enrichment of both standard and prototype is the same: 20% U 235 and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs

RA Reactor operation and maintenance (I-IX), part VII, Task 3.08/04, Refurbishment of the RA reactor

International Nuclear Information System (INIS)

Zecevic, V.

1963-12-01

This volume covers the following reports concerned with the maintenance and repair work of the RA reactor: repair of the technical water system; maintenance of the transportation equipment; vacuuming and drying during refurbishment; repair and decontamination of the distillation device; and the report on participation of the operational dosimetry division in the RA reactor refurbishment activities

Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

International Nuclear Information System (INIS)

Park, Cheol; Lee, B. C.; Chae, H. T. and others

2006-01-01

A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor

3 MW TRIGA Research Reactor facility of BAEC and its Utilization

Energy Technology Data Exchange (ETDEWEB)

Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z. [Bangladesh Atomic Energy Commission, Atomic Energy Research Establishment, Dhaka (Bangladesh)

1999-08-01

The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

Modular 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms

Science.gov (United States)

Podwin, Agnieszka; Dziuban, Jan A.

2017-10-01

The paper presents the sandwiched polymer 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms. Euglenas and yeast were separately and simultaneously cultured for 10 d in the chip. As a result of the experiments, euglenas, light-initialized and nourished by CO2—a product of ethanol fermentation handled by yeast—generated oxygen, based on the photosynthesis process. The presence of oxygen in the bio-reactor was confirmed by the colorimetric method—a bicarbonate (pH) indicator. Preliminary studies towards the obtainment of an effective source of oxygen are promising and further research should be done to enable the utility of the bio-reactor in, for instance, microbial fuel cells.

Modular 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms

International Nuclear Information System (INIS)

Podwin, Agnieszka; Dziuban, Jan A

2017-01-01

The paper presents the sandwiched polymer 3D printed lab-on-a-chip bio-reactor for the biochemical energy cascade of microorganisms. Euglenas and yeast were separately and simultaneously cultured for 10 d in the chip. As a result of the experiments, euglenas, light-initialized and nourished by CO 2 —a product of ethanol fermentation handled by yeast—generated oxygen, based on the photosynthesis process. The presence of oxygen in the bio-reactor was confirmed by the colorimetric method—a bicarbonate (pH) indicator. Preliminary studies towards the obtainment of an effective source of oxygen are promising and further research should be done to enable the utility of the bio-reactor in, for instance, microbial fuel cells. (paper)

ASN’s actions in GEN IV reactors and Sodium Fast Reactors (SFR)

International Nuclear Information System (INIS)

Belot, Clotilde

2013-01-01

The ASN is involved in 3 actions concerning GEN IV: • Overview of nuclear reactor GEN IV systems; • Specific analysis about transmutation; • Prototype reactor ASTRID (SFR). Furthermore theses actions are in the beginning (no conclusions or results available)

Mutational characterization of the P3H1/CRTAP/CypB complex in recessive osteogenesis imperfecta.

Science.gov (United States)

Barbirato, C; Trancozo, M; Almeida, M G; Almeida, L S; Santos, T O; Duarte, J C G; Rebouças, M R G O; Sipolatti, V; Nunes, V R R; Paula, F

2015-12-03

Osteogenesis imperfecta (OI) is a genetic disease characterized by bone deformities and fractures. Most cases are caused by autosomal dominant mutations in the type I collagen genes COL1A1 and COL1A2; however, an increasing number of recessive mutations in other genes have been reported. The LEPRE1, CRTAP, and PPIB genes encode proteins that form the P3H1/CRTAP/CypB complex, which is responsible for posttranslational modifications of type I collagen. In general, mutations in these genes lead to severe and lethal phenotypes of recessive OI. Here, we describe sixteen genetic variations detected in LEPRE1, CRTAP, and PPIB from 25 Brazilian patients with OI. Samples were screened for mutations on single-strand conformation polymorphism gels and variants were determined by automated sequencing. Seven variants were detected in patients but were absent in control samples. LEPRE1 contained the highest number of variants, including the previously described West African allele (c.1080+1G>T) found in one patient with severe OI as well as a previously undescribed p.Trp675Leu change that is predicted to be disease causing. In CRTAP, one patient carried the c.558A>G homozygous mutation, predicted as disease causing through alteration of a splice site. Genetic variations detected in the PPIB gene are probably not pathogenic due to their localization or because of their synonymous effect. This study enhances our knowledge about the mutational pattern of the LEPRE1, CRTAP, and PPIB genes. In addition, the results strengthen the proposition that LEPRE1 should be the first gene analyzed in mutation detection studies in patients with recessive OI.

Review of mirror fusion reactor designs

International Nuclear Information System (INIS)

Bender, D.J.

1977-01-01

Three magnetic confinement concepts, based on the mirror principle, are described. These mirror concepts are summarized as follows: (1) fusion-fission hybrid reactor, (2) tandem mirror reactor, and (3) reversed field mirror reactor

Comparison of 'system thermal-hydraulics-3 dimensional reactor kinetics' coupled calculations using the MARS 1D and 3D modules and the MASTER code

International Nuclear Information System (INIS)

Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.

2002-01-01

KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations

The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

International Nuclear Information System (INIS)

Brand, P.; Wiechers, W.K.

1974-01-01

The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

Cost analysis and financial risk profile for severe reactor accidents at Waterford-3

International Nuclear Information System (INIS)

Cutbush, J.D.; Abbott, E.C.; Carpenter, W.L. Jr.

1992-01-01

To support Louisiana Power and Light Company (LP and L) in determining an appropriate level of nuclear property insurance for Waterford Steam Electric Station, Unit 3 (Waterford-3), ABZ, Incorporated, performed a series of cost analyses and developed a financial risk profile. This five-month study, conducted in 1991, identified the potential Waterford-3 severe reactor accidents and described each from a cleanup perspective, estimated the cost and schedule to cleanup from each accident, developed a probability distribution of associated financial exposure, and developed a profile of financial risk as a function of insurance coverage

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

International Nuclear Information System (INIS)

Huda, M.Q.

2006-01-01

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

Differential effects of collagen prolyl 3-hydroxylation on skeletal tissues.

Directory of Open Access Journals (Sweden)

Erica P Homan

2014-01-01

Full Text Available Mutations in the genes encoding cartilage associated protein (CRTAP and prolyl 3-hydroxylase 1 (P3H1 encoded by LEPRE1 were the first identified causes of recessive Osteogenesis Imperfecta (OI. These proteins, together with cyclophilin B (encoded by PPIB, form a complex that 3-hydroxylates a single proline residue on the α1(I chain (Pro986 and has cis/trans isomerase (PPIase activity essential for proper collagen folding. Recent data suggest that prolyl 3-hydroxylation of Pro986 is not required for the structural stability of collagen; however, the absence of this post-translational modification may disrupt protein-protein interactions integral for proper collagen folding and lead to collagen over-modification. P3H1 and CRTAP stabilize each other and absence of one results in degradation of the other. Hence, hypomorphic or loss of function mutations of either gene cause loss of the whole complex and its associated functions. The relative contribution of losing this complex's 3-hydroxylation versus PPIase and collagen chaperone activities to the phenotype of recessive OI is unknown. To distinguish between these functions, we generated knock-in mice carrying a single amino acid substitution in the catalytic site of P3h1 (Lepre1(H662A . This substitution abolished P3h1 activity but retained ability to form a complex with Crtap and thus the collagen chaperone function. Knock-in mice showed absence of prolyl 3-hydroxylation at Pro986 of the α1(I and α1(II collagen chains but no significant over-modification at other collagen residues. They were normal in appearance, had no growth defects and normal cartilage growth plate histology but showed decreased trabecular bone mass. This new mouse model recapitulates elements of the bone phenotype of OI but not the cartilage and growth phenotypes caused by loss of the prolyl 3-hydroxylation complex. Our observations suggest differential tissue consequences due to selective inactivation of P3H1 hydroxylase

Physics analysis of the Apollo D-3He tokamak reactor

International Nuclear Information System (INIS)

Santarius, J.F.; Emmert, G.A.

1990-01-01

Recent developments in the analysis and conceptual design of Apollo, a D- 3 He Tokamak Reactor are presented. Encouraging experimental results on TEXT motivated a key change in the Apollo concept utilization of an ergodic magnetic limiter for impurity control instead of a divertor. Parameters for the updated Apollo design and an analysis of the ergoidc magnetic limiter are given. The Apollo reference case uses direct conversion of synchrotron radiation to electricity by rectifying antennas (rectennas) for its power conversion system. Previous analyses of this concept are expanded, including further details of the rectennas and of the loss of synchrotron power to the waveguides and walls. Although Apollo will burn D- 3 He fuel, a significant amount of unburned tritium will be generated by D4D reactions. The possibility of operating a short, dedicated, T+ 3 He burn phase to eliminate this tritium will be examined

A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

International Nuclear Information System (INIS)

Dubourg, M.

1993-01-01

This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

Repairing liner of the reactor; Reparacion del liner del reactor

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-15

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

Reactor Structure Materials: Corrosion of Reactor Core Internals

International Nuclear Information System (INIS)

Van Dyck, S.

2000-01-01

The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

Principle tests of reactor pumped laser of 3He-Ar-Xe system

International Nuclear Information System (INIS)

Chen Hande; Jin Xingxing; Yang Chengde

1994-01-01

A nuclear reactor-pumped laser was demonstrated firstly by using the 3 He(P,n) 3 H reaction to excite a 3 He-Ar-Xe laser. Lasing was observed on the 1.73 μm (5d (3/2) 1 -6p(5/2) 1 transition) in Xe I. The CAEP Pulsed Reactor (CFBR-II) was used as a source of fast neutrons moderated by a 50 mm thick polyethylene sleeve around the laser cell. A laser cell was constructed from K 9 glass tubing 600 mm long and Φ34 mm with each end cut at Brewster's angle (quartz). The laser cell was baked out to 10 -3 Pa prior to filling with research-grade mixture ( 3 He/Ar/Xe = 34.7: 34.7: 0.267 kPa). A dielectric-coated 2 m radius-curvature back mirror (99.7% reflectivity at 1.73 μm) and a flat output coupler (Φ30 mm) were used to form the optical cavity. The cavity optics were placed outside the cell as shown in Fig.3. A schematic of the experimental configuration is shown. Typical laser signals obtained with the Ge detector (-80 degree C) are also shown. Laser output lags the fast neutron pulse by the neutron thermalization time (∼278 μs). The observed light output was shown to be caused by stimulated emission, because the signal disappeared when the reflecting (flat) mirror was covered. This precludes the possibility that the phenomenon was the result of γ-radiation interaction with the detector or the air in the light path. The measured laser energy was 80 times the maximum possible fluorescence energy. The observed wavelength was measured to be 1.73 μm by using glass filters. Laser output duration was approximately 735 μs

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)

Report on the Survey of the Design Review of New Reactor Applications. Volume 3: Reactor

International Nuclear Information System (INIS)

Downey, Steven; Monninger, John; Nevalainen, Janne; Lorin, Aurelie; ); Webster, Philip; Joyer, Philippe; Kawamura, Tomonori; Lankin, Mikhail; Kubanyi, Jozef; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Kang, Kyungmin; Kim, Walter

2016-01-01

At the tenth meeting of the 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 document, which is the third report on the results of the Design Phase Survey, focuses on the Reactor. The Reactor category includes the following technical topics: fuel system design, reactor internals and core support, nuclear design and core nuclear performance, thermal and hydraulic design, reactor materials, and functional design of reactivity control 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. - Design review strategies most commonly used to confirm that the regulatory requirements have been met include document review and independent verification of calculations, computer codes, or models used to describe the design and performance of the core and the fuel. - It is common to consider operating experience and

Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964 - Annex 12a

International Nuclear Information System (INIS)

Pavicevic, M.

1964-01-01

Conclusions of the experts group of the RA reactor are related to: analyses of reactor operation at 6.5 MW power with heavy water coolant flow of 250 m 3 /h (2 pumps rotation speed 1500 rotations/min); decisions of future operation; further preparation activities related to reactor operation in forced regime and reduced cooling conditions

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

Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

International Nuclear Information System (INIS)

Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

2014-01-01

This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

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

A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

Energy Technology Data Exchange (ETDEWEB)

Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

2003-07-01

Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

3-dimensional finite element modelling of reactor building internal structure for static analysis

International Nuclear Information System (INIS)

Joshi, M.H.; Reddy, V.J.; Kushwaha, H.S.; Reddy, G.R.; Karandikar, G.V.

1991-01-01

a) Thin shell element gives fairly accurate results when compared to 3-D Brick element for the type of structure and loading in Reactor Building. b) The maximum element size is fixed from model 3(c) i.e. 2.0 m. c) Openings with size smaller than 0.5 m can be neglected without affecting the results very much. d) For any such problem, the methodology described in this paper can be used to take rational decisions which will ensure reasonable accuracy. (author)

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

Nuclear reactor

International Nuclear Information System (INIS)

Schulze, I.; Gutscher, E.

1980-01-01

The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

Pressurized water reactors: the EPR project

International Nuclear Information System (INIS)

Py, J.P.; Yvon, M.

2007-01-01

EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

Power measurement of the RA-3 reactor using the neutron noise technique and 16N

International Nuclear Information System (INIS)

Gomez, Angel

2003-01-01

This work describes a measurement method based on the neutron noise technique which is used for determining the relation between the power and the currents of two ionization chambers. These chambers are sensitive to the gamma radiation from the 16 N decay produced in the RA-3 reactor core. The power during operation is obtained from the calibration factors by measuring those currents. As this calibration factors depend on the cooler flow that circulates in the reactor core and in the 16 N measuring system, an estimator, that is a function of the ratio of this currents, is proposed in order to detect flow changes. (author)

Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

Energy Technology Data Exchange (ETDEWEB)

Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

2015-09-15

This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

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.)

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

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

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

Optimization of reactor power by taking into consideration temperature increase in a reactor pumped 3He-Xe laser

International Nuclear Information System (INIS)

Cetin, Fuesun

2009-01-01

In nuclear pumped lasers, gas parameters are optimized in a manner such that output power is increased for constructing a high power laser. Since output power increases with the increase of energy deposited in the gas, high output power requires high pumping power. However, the high energy loading results in elevated gas temperature. Temperature increase of this magnitude can detrimentally influence the laser gain and efficiency, since it negatively impacts several important laser kinetic.processes. This fact may cause laser output to abruptly terminate before the peak of the pump pulse [1-3]. A nuclear pumped laser using a volumetric energy source through the 3 He(n, p) 3 H reaction has here been considered. It is assumed that TRIGA Mark II Reactor at Istanbul Technical University is used for nuclear pumping as the neutron source. In the previous papers, the optimal parameters for improving both output power and optical homogeneity were determined [4-5]. Spatial and temporal variations of gas temperature during pumping pulse for maximum peak power (1200 MW) were determined for various operating pressures in Ref. [6]. It was seen that gas temperature reaches up to 1000 0 K near the peak of the pumping pulse for the initial pressures of 1-4 atm. This means that laser output may terminate before the peak of the pump pulse due to overheating of laser gas. Under these conditions, a question arises about a further optimisation taking into consideration gas temperature. This question has been examined in this study. Experimental results (Batyrbekov et al, 1989) showed that temperature rise up to 650 C had no influence on Xe laser characteristics [ 7]. Therefore, It has here been assumed that the lasing will terminate when gas temperature reaches 1000 0 K for a Xe-laser with 3 He buffer gas. Under these conditions optimum reactor power is investigated by taking into consideration lasing duration also. (orig.)

Semi-catalyzed deuterium reactors for co-generation of 3He and synfuels (the CoSCD concept)

International Nuclear Information System (INIS)

1980-01-01

The potential of developing semi-catalyzed deuterium reactors for co-generation of 3 He and synthetic fuels is discussed. Such factors as environmental impact, siting, energy basics, and engineering technology are also discussed

Development of Pneumatic Transfer Irradiation Facility (PTS no.3) for Neutron Activation Analysis at HANARO Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

2008-04-15

A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

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

1991-06-01

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

Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0